[Federal Register Volume 64, Number 230 (Wednesday, December 1, 1999)]
[Notices]
[Pages 67330-67348]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-31037]


-----------------------------------------------------------------------

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) 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 November 6, 1999, through November 19, 1999. 
The last biweekly notice was published on November 17, 1999 (64 FR 
62704).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications 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 January 3, 2000, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and

[[Page 67331]]

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.
    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: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. 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).

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

    Date of amendment request: October 12, 1999.
    Description of amendment request: This proposed technical 
specification change removes the anticipatory reactor scram signal for 
turbine electro-hydraulic control (EHC) low oil pressure trip from the 
reactor protection system (RPS) trip function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Does the change involve a significant increase in the probability 
of occurrence or consequences of an accident previously evaluated?

    The proposed change removes the Turbine EHC Control Oil 
Pressure-Low scram function and the associated Limiting Safety 
System Setting (LSSS). The purpose of the Turbine EHC Control Oil 
Pressure scram is to anticipate the pressure transient which would 
be caused by imminent control valve closure on loss of control oil 
pressure. This

[[Page 67332]]

function does not serve as an initiator for any accidents evaluated 
in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). 
In addition, this trip function is not credited in any design basis 
event and is functionally redundant to the Turbine Control Valve 
Fast Closure RPS trip function during a postulated loss of EHC 
control oil event. The Turbine Control Valve Fast Closure will 
initiate a scram on a loss of control oil event coincident with 
turbine control valve closure.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The removal of this function does not represent a change in 
operating parameters or introduce a new mode of operation. The 
pressure switches associated with the Turbine Control Valve Fast 
Closure function provide equivalent protection from a loss of EHC 
oil event. For this reason, the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    Operation under the proposed amendment will not change any plant 
operation parameters, nor any protective system actuation setpoints 
other than removal of the Turbine EHC Control Oil Pressure-Low scram 
function. The scram function associated with the Turbine Control 
Valve Fast Closure provides equivalent protection for events 
involving fast turbine control valve closure including the loss of 
EHC control oil pressure. For this reason, eliminating the EHC 
Control Oil Pressure-Low scram function, which is redundant to other 
protective instrumentation, does not reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: September 23, 1999.
    Description of amendment request: The proposed amendment would 
relocate items associated with instrumentation for toxic gas monitoring 
from the Technical Specifications (TSs) to the Updated Final Safety 
Analysis Report (UFSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve a significant hazards 
consideration because:

    1. There is no significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are administrative in nature. The 
Specifications and associated Bases will be transferred verbatim to 
the UFSAR.
    These changes do not affect possible initiating events for 
accidents previously evaluated or alter the configuration or 
operating of the facility. The Limiting Safety Systems Settings and 
Safety Limits specified in the current TSs remain unchanged. 
Therefore, the proposed changes to the subject TS would not increase 
the probability or consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any accident previously evaluated has not been created.
    As stated above, the proposed changes are administrative in 
nature. The safety analysis of the facility remains complete and 
accurate. There are no physical changes to the facility, and the 
plant conditions for which the design basis accidents have been 
evaluated are still valid. The operating procedures and emergency 
procedures are unaffected. Consequently, no new failure modes are 
introduced as a result of the proposed changes, therefore, the 
proposed changes will not initiate any new or different kind of 
accident.
    3. There has been no significant reduction in the margin of 
safety.
    The proposed changes are administrative in nature. Since there 
are no changes to the operation of the facility or physical design, 
the UFSAR design basis, accident assumptions are not affected. 
Therefore, the proposed changes will not result in a reduction in 
the margin of safety.
    The proposed changes have been reviewed by both the Station 
Nuclear Safety Committee (SNSC) and the Con Edison Nuclear Facility 
Safety Committee (NFSC). Both Committees concur that the proposed 
changes do not represent a significant hazards consideration.

    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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 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 amendment request: November 3, 1999.
    Description of amendment request: The amendments would revise 
Section 3.8.1, ``AC [alternating current] Sources--Operating,'' of the 
Technical Specifications. Specifically, this would revise: (1) 
Surveillance Requirement (SR) 3.8.1.9 to delete the power factor 
requirement from the diesel generator (DG) load rejection test; (2) SR 
3.8.1.13 to allow performance of the diesel generator non-emergency 
automatic trip bypass test at any operational power level; and (3) SR 
3.8.1.14 to allow performance of the 24-hour diesel generator run at 
any operational power level and delete the power factor requirement. No 
plant modification is involved with this proposed amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a reduction in a margin of safety.

First Standard

    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 effect 
on accident probabilities or consequences. The DGs and their 
associated emergency buses are not accident initiating equipment; 
therefore, there will be no impact on any accident probabilities by 
the approval of this amendment. The design of the equipment is not 
being modified by these proposed changes. In addition, the ability 
of the DGs to respond to a design basis accident will not be 
adversely impacted by these proposed changes. There will be no 
significant increased likelihood of causing a blackout of a safety 
bus by the proposed changes in testing. Therefore, there will be no 
significant impact on any accident consequences.

Second Standard

    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. Equipment 
will be operated in the same configuration with the exception of the 
plant

[[Page 67333]]

mode in which the testing is conducted. No changes are being made to 
the plant which will introduce any new accident causal mechanisms. 
This amendment request does not impact any plant systems that are 
accident initiators; neither does it adversely impact any accident 
mitigating systems.

Third Standard

    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. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed amendment. The equipment referenced in the revised TS for 
these proposed changes is already capable of performing as designed. 
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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    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: November 3, 1999.
    Description of amendment request: The proposed amendments would 
revise Section 3.8.1, ``AC [alternating current] Sources--Operating,'' 
of the Technical Specifications. Specifically, this would revise: (1) 
Surveillance Requirement (SR) 3.8.1.9 to allow performance of the 
diesel generator (DG) load rejection test at any operational power 
level and to delete the power factor requirement; (2) SR 3.8.1.10 to 
allow performance of the diesel generator full load rejection test at 
any operational power level; and (3) SR 3.8.1.14 to allow performance 
of the 24-hour diesel generator run at any operational power level and 
delete the power factor requirement. No plant modification is involved 
with this proposed amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, 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.

First Standard

    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 effect 
on accident probabilities or consequences. The DGs and their 
associated emergency buses are not accident initiating equipment; 
therefore, there will be no impact on any accident probabilities by 
the approval of this amendment. The design of the equipment is not 
being modified by these proposed changes. In addition, the ability 
of the DGs to respond to a design basis accident will not be 
adversely impacted by these proposed changes. There will be no 
significant increased likelihood of causing a blackout of a safety 
bus by the proposed changes in testing. Therefore, there will be no 
significant impact on any accident consequences.

Second Standard

    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. Equipment 
will be operated in the same configuration with the exception of the 
plant mode in which the testing is conducted. No changes are being 
made to the plant which will introduce any new accident causal 
mechanisms. This amendment request does not impact any plant systems 
that are accident initiators; neither does it adversely impact any 
accident mitigating systems.

Third Standard

    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. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed amendment. The equipment referenced in the revised TS for 
these proposed changes is already capable of performing as designed. 
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: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: October 7, 1999.
    Description of amendment request: Grand Gulf Nuclear Station (GGNS) 
requests approval to revise its licensing basis for the release of 
fission products following an accident. The basis for the proposed 
change makes use of one of the insights established in NUREG-1465, 
``Accident Source Terms for Light Water Nuclear Power Plants,'' which 
defines alternative source terms for use in the licensing of light 
water reactors. Specifically, this application credits the insight that 
there is a delay in the release of fission products from the reactor 
fuel following a postulated design basis loss-of-coolant accident 
(LOCA). The timing of fission product release from fuel perforation, 
i.e., gap activity release, is based on the boiling water reactor 
(BWR)--specific value of the timing of the gap activity release phase 
of a LOCA as calculated in the Boiling Water Reactor Owners Group 
(BWROG) Report, ``Prediction of the Onset of Fission Gas Release From 
Fuel in Generic BWR.'' This BWROG Report has been previously reviewed 
and approved by the Nuclear Regulatory Commission (NRC) staff. The 
licensing basis change to Updated Final Safety Analysis Report (UFSAR) 
Section 15.6.5.5.2 proposed by GGNS replaces the assumption of an 
instantaneous release of gap activity phase fission products into the 
drywell with a more accurate scenario in which the gap activity release 
is delayed by up to 121 seconds as calculated in the BWROG Report. 
Approval of this change will allow GGNS to increase the containment 
isolation valve closure times credited for limiting post-accident doses 
to both control room personnel and to offsite individuals. While this 
new basis would be applicable to all of the containment isolation 
valves, it addresses only the dose mitigation aspects of the closure 
requirements. There are currently some valves for which the closure 
time is limited based on other functional performance requirements 
(e.g., line break isolation). This submittal does not propose any 
changes that would

[[Page 67334]]

eliminate any of these other requirements. The allowable closure times 
for these valves would not be affected by this proposed change.
    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:

    GGNS staff has evaluated the proposed change to incorporate a 
delay in the post-accident fission product release into its 
licensing basis. This change recognizes one of the revised source 
term insights discussed in NUREG-1465. This change in the licensing 
basis will provide the basis for revising the Technical Requirements 
Manual to increase Primary Containment Isolation Valve (PCIV) 
maximum isolation times. These changes have been evaluated using the 
standards in 10CFR50.92 and it is concluded that they do not involve 
any significant hazards considerations. Specifically, the proposed 
change will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    The proposed change takes credit for a new source term insight 
that recognizes that the fission product release from a fuel 
assembly is not instantaneous with a design basis accident. 
Implementation of this change into the licensing basis will be used 
to justify an increase in the maximum allowable PCIV isolation 
times. These changes do not affect the precursors for any accident 
or transient evaluated in Chapter 15 of the GGNS UFSAR. Therefore, 
there is no increase in the probability of any accident previously 
evaluated.
    A plant specific radiological analysis has been performed to 
evaluate the effect on the dose consequences of extending the 
maximum allowable closure time. This evaluation considered the 
initial two-minute period of the accident during which, according to 
new source term insights developed in NUREG-1465 and in a BWROG 
report, fission product releases are not expected to occur. Releases 
from the break and from containment during this period consist of 
coolant radioactivity only. The total release during this period was 
found to result in an offsite dose of less than 0.60 rem. This dose 
represents only a small fraction of the LOCA dose evaluated in the 
UFSAR. As this submittal is for a limited scope application of the 
NUREG-1465 insights (in this case, timing and duration of the 
coolant activity phase) and addresses only the first 121 seconds of 
the accident scenario, the total long-term dose determined using the 
TID-14844 assumptions is not changed by this submittal.
    In reality, the other insights offered in the NUREG would be 
expected to result in an overall dose reduction. In any event, the 
dose consequences of the proposed change do not result in an 
increase in the consequences of any accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated;
    The primary containment isolation system is designed to prevent, 
as much as practicable, the unfiltered release of radioactive 
material to the environs following an accident. As such, the system 
is relied upon for accident dose consequence mitigation. Neither the 
revision of the licensing basis to recognize that fission product 
releases are not instantaneous as is assumed in the current 
analysis, nor the extension of the valve closure times affects the 
ability of the valves to perform their accident mitigation function. 
It is also noted that the increased closure time allowables will 
only be applied to valves which do not have an alternate 
constraining performance requirement for closure time; the safety 
functions of other supported components and systems are not 
affected. Thus, the proposed change does not create the potential 
for a new or different kind of accident.
    (3) Involve a significant reduction in a margin of safety.
    The proposed change revises the bases for the offsite dose 
calculation to credit, in the initial 2 minutes of the accident 
scenario, the fact that there is no fuel failure expected during 
this time. That is, for the first two minutes of the event, only 
coolant activity is released. The other assumptions, bases and 
methodologies for offsite dose calculations used to evaluate the 
long-term offsite dose consequences of accidents described in FSAR 
[Final Safety Analysis Report] Chapter 15 are not affected by this 
change. The margin between calculated dose consequences described in 
the FSAR and regulatory limits is not reduced.
    A recent GGNS analysis of the LOCA scenario considering the only 
release in the first 121 seconds is from the reactor coolant 
resulted in an EAB [exclusion area boundary] dose of less than 1 rem 
thyroid during this period. The total dose for the 0- to 2-hour 
period is not expected to increase due to the delay in the fission 
product release; the total amount of radioactivity released will 
remain the same. Both the recently evaluated 2-minute dose and the 
24.9 rem in two hours as presented in the UFSAR are insignificant in 
comparison to the 300 rem acceptance limit for this scenario. The 
GGNS SER [safety evaluation report] acknowledges the conservatism of 
the old analysis methodology. An independent analysis done by the 
staff during their evaluation of the GGNS FSAR estimated doses could 
decrease about 95% if the fission product release were to be delayed 
by 2 minutes.
    The bases for PCIV closure times described in the Technical 
Specifications remain unchanged. The inconsistency between the 
assumption of immediate containment isolation in the dose analysis 
and allowable isolation valve closure times of one to two minutes is 
eliminated by this change. Plant specific analysis has shown that 
the expected dose resulting from the PCIVs remaining open during 
this period is insignificant.
    Actual safety benefits are expected to result from valve 
performance and reliability improvements, elimination of unnecessary 
reports and system performance improvements such as minimization of 
water hammer events. Therefore, the increase in maximum isolation 
time for certain PCIVs proposed in this submittal will not result in 
a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit No. 1, Dauphin County, Pennsylvania.

    Date of amendment request: August 20, 1999.
    Description of amendment request: The proposed license amendment 
would modify the Technical Specifications (TSs) to allow revision of 
the 4KV Engineered Safeguards Bus Undervoltage Relay Degraded Voltage 
calibration to be performed at an annual interval rather than its 
present refueling interval and change the bases to state that the 
degraded voltage relay setpoint tolerance is being changed from an ``as 
left'' reading to an ``as found'' reading. Additionally, the new 
calculations supporting the request identified a need to compensate for 
lack of voltage margin through reliance on manual action in lieu of 
full automatic voltage protection, as implied by Chapter 8 of the 
Updated Final Safety Analysis Report (UFSAR). Such actions would 
involve load manipulations following a loss of coolant accident (LOCA) 
with post LOCA conditions in combination with extremely low switchyard 
voltage. An additional limit of operation with a maximum of 5 
Circulating Water pumps while in single 230KV auxiliary transformer 
operation is also added to the UFSAR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes to the degraded voltage relay setpoint 
tolerance and calibration interval are intended to reduce the total 
degraded voltage relay setpoint uncertainties. These changes will 
provide greater confidence that minimum voltages necessary to 
operate NSR [nuclear safety related] equipment are not exceeded. In 
combination, the proposed changes for degraded voltage relay 
setpoint tolerance and

[[Page 67335]]

calibration interval will reduce the probability that ES [engineered 
safeguards] buses will be separated from their offsite power source 
during low grid voltage conditions. This will reduce challenges to 
the onsite emergency power systems. The proposed changes will 
enhance the ability of the undervoltage protection scheme to perform 
in accordance with its intended design, and will improve the ability 
of the scheme to respond to low voltage conditions caused by 
malfunction of equipment important to safety.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated in the SAR.
    2. The proposed setpoint tolerance and calibration interval 
changes are consistent with the specifications and intended design 
of the degraded voltage protection scheme and do not introduce the 
possibility of any new failure modes to the protection scheme or the 
electrical distribution system. The proposed changes reduce the 
probability of insufficient voltage to NSR loads and reduce the 
probability of separation of ES buses from the offsite power source. 
Therefore, operation of the facility in accordance with the proposed 
changes do not create a possibility of a new or different type of 
accident than any previously evaluated in the SAR.
    3. The proposed setpoint tolerance and calibration interval 
changes are intended to reduce the total degraded voltage relay 
setpoint uncertainties. The changes will provide greater confidence 
that minimum voltages necessary to operate NSR equipment will not be 
exceeded. The proposed changes will also reduce the probability that 
the ES buses will be separated from their offsite power source 
during low grid voltage conditions. These effects will enhance the 
objective [of] providing a reliable source of power for BOP 
auxiliaries and [a] continuously available power supply for the ES 
equipment as required by TS [technical specification] 3.7 bases. 
Therefore, operation of the facility in accordance with the proposed 
changes would not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that 
the amendment request involves no significant hazards consideration.

    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment requests: November 3, 1999.
    Description of amendment requests: The proposed amendments would 
allow use of fuel rods with ZIRLO cladding, specify an alternate 
methodology to determine the integral fuel burnable absorber (IFBA) 
requirements for Westinghouse fuel assemblies stored in the new fuel 
storage racks, and delete the designation of the fuel assembly types 
allowed in the spent fuel storage racks and the new fuel storage racks.
    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 of occurrence or consequences of an accident previously 
evaluated?
    The proposed T/S [Technical Specification] change to allow 
storage and use of fuel rods clad with ZIRLO does not significantly 
increase the probability of occurrence of an accident. Fuel 
assemblies are not an initiator or precursor to any previously 
evaluated accident. The proposed T/S change does not change or alter 
the design criteria for the systems or components used to mitigate 
the consequences of any design basis accident. Use of ZIRLO fuel 
cladding does not adversely affect fuel performance or impact 
nuclear design methodology. Therefore, accident analysis results are 
not impacted. The operating limits are not changed and the analysis 
methods to demonstrate operation within the limits remain in 
accordance with NRC-approved methodologies. Other than the changes 
to the fuel rod cladding there are no physical changes to the plant 
associated with this T/S change. A safety analysis is still required 
to be performed for each specific reload cycle to demonstrate 
compliance with fuel safety design bases. The 10 CFR 50.46 emergency 
core cooling system acceptance criteria are applied to the ZIRLO 
clad fuel rods. The use of fuel assemblies containing ZIRLO clad 
fuel rods does not result in a change to the reload design and 
safety analysis limits. The clad material is similar in chemical 
composition and has similar physical and mechanical properties as 
Zircaloy-4. Thus, the cladding integrity is maintained and the 
structural integrity of the fuel assembly is not affected. ZIRLO 
cladding improves corrosion performance and dimensional stability. 
Since the dose predictions in the safety analyses are not sensitive 
to the fuel rod cladding material used, the radiological 
consequences of accidents previously evaluated in the safety 
analysis remain valid.
    The proposed T/S change to specify an alternate NRC-approved 
methodology used to determine the IFBA requirements for Westinghouse 
fuel assemblies stored in the new fuel storage racks does not change 
or alter the design criteria for the systems or components used to 
mitigate the consequences of any design basis accident. This 
alternate methodology is more conservative with respect to 
determining the reactivity of the stored fuel assemblies than the 
methodology currently specified in the T/S. Therefore, the 
probability of an accidental criticality is less with the proposed 
T/S change than currently assumed. Since a criticality accident is 
precluded by the proposed T/S change, the consequences of a 
criticality accident are not changed by the use of this alternate 
methodology.
    The proposed T/S change to delete designation of the fuel 
assembly types allowed in the spent fuel storage racks and new fuel 
storage racks is administrative, and does not alter the design and 
analysis requirements that ensure storage of fuel in safe 
configurations. The existing T/S requirements for maximum 
enrichment, reactivity, and spacing of fuel assemblies in the spent 
fuel storage racks and new fuel storage racks are not altered by 
this change.
    Based on the above discussions, design basis accident analyses 
affected by these 
T/S changes remain valid, and the consequences of an accident 
previously evaluated are not significantly increased by these 
changes.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed T/S change to allow storage and use of fuel rods 
clad with ZIRLO cannot create a new or different kind of accident. 
Fuel assemblies with ZIRLO clad fuel rods satisfy the same design 
bases as those used for fuel assemblies with Zircaloy-4 clad fuel 
rods. The design and performance criteria continue to be met and no 
new failure mechanisms have been identified. Since the original 
design criteria are met, the ZIRLO clad fuel rods cannot be an 
initiator for any new accident. The ZIRLO cladding material offers 
improved corrosion resistance and structural integrity. The proposed 
changes do not affect the design or operation of any other system or 
component in the plant. The safety functions of the other 
structures, systems, or components are not changed in any manner, 
nor is the reliability of any other structure, system, or component 
reduced. The changes do not affect the manner by which the facility 
is operated and do not change any other facility design feature, 
structure, or system. No new or different types of permanent plant 
equipment are installed by this proposed 
T/S change. In addition, the use of ZIRLO fuel assemblies does not 
involve any alterations to permanent plant equipment or plant 
operating procedures that would introduce any new or unique 
operational mode or accident precursor.
    The proposed T/S change to specify an alternate NRC-approved 
methodology used to determine the IFBA requirements for Westinghouse 
fuel assemblies stored in the new fuel storage racks ensures that a 
conservative methodology is used to verify the licensing basis 
reactivity limits are not exceeded. The proposed change does not 
affect any permanent plant equipment or plant operating procedures, 
and cannot be an initiator of an event.
    The proposed T/S change to delete designation of the fuel 
assembly types allowed in the spent fuel storage racks and new fuel 
storage racks is an administrative

[[Page 67336]]

change only. The proposed change does not affect any permanent plant 
equipment or plant operating procedures, and cannot be an initiator 
of an event.
    Since there is no change to the permanent facility or plant 
operating procedures, and the safety functions and reliability of 
structures, systems, or components are not affected, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, it is concluded that the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed T/S change to allow storage and use of fuel rods 
clad with ZIRLO does not change the reactor fuel reload design and 
safety analysis limits. The use of these fuel assemblies takes into 
consideration the core operating conditions allowed in the T/S. For 
each cycle reload core, the fuel assembly design and core 
configuration are evaluated using NRC-approved reload design 
methods, including consideration of the core physics analysis 
peaking factors and core average linear heat rate effects. The 
design basis and modeling techniques for fuel assemblies with 
Zircaloy-4 clad fuel rods remain valid for fuel assemblies with 
ZIRLO clad fuel rods. Use of ZIRLO cladding material has no effect 
on the criticality analysis for the spent fuel storage racks and the 
new fuel storage racks. Furthermore, it has no effect on the 
thermal-hydraulic and structural analysis for the spent fuel pool. 
Therefore, the design and safety analysis limits specified in the T/
S are maintained with this proposed change.
    The proposed T/S change to specify an alternate NRC-approved 
methodology used to determine the IFBA requirements for Westinghouse 
fuel assemblies stored in the new fuel storage racks ensures that a 
conservative methodology is used to verify the licensing basis 
reactivity limits are not exceeded. Therefore, the existing T/S 
margin for reactivity control in the new fuel storage racks is 
maintained by this proposed change.
    The proposed T/S change to delete designation of the fuel 
assembly types allowed in the spent fuel storage racks and new fuel 
storage racks is an administrative change, and does not alter any of 
the existing T/S limits governing storage and use of reactor fuel.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: October 16, 1998, as supplemented by 
letters dated December 30, 1998, May 10, June 15, July 30, August 2, 
11, 16, 19, 27, September 10, and 30, 1999.
    Description of amendment request: Associated with a Niagara Mohawk 
Power Corporation (NMPC or the licensee) application to convert from 
the Curent Technical Specifications (CTS) for the Nine Mile Point 
Nuclear Power Station, Unit No. 2, to Improved Technical Specifications 
(ITS) as contained in Revision 1 of NUREG-1433, and Revision I of 
NUREG-1434, ``Standard Technical Specifications for General Electric 
Plants, BWR/4 and BWR/6'' dated April 1995, the licensee proposed to 
allow two hydrogen recombiners to be inoperable for up to 7 days 
provided that the alternate hydrogen control system is found to be 
acceptable to the NRC staff as described below.
    CTS 3.6.6.1 ACTION only permits one hydrogen recombiner to be 
inoperable. If two hydrogen recombiners are inoperable, CTS 3.0.3 is 
entered. CTS 3.6.6.1 ACTION has been modified to incorporate Standard 
Technical Specification (STS) 3.6.3.1 ACTION B which allows two 
hydrogen recombiners to be inoperable for up to 7 days. The use of STS 
3.6.3.1 ACTION B is allowed, as specified in a Bases Reviewer's Note, 
provided that the alternate hydrogen control system is found to be 
acceptable to the NRC staff. Therefore, the licensee proposed to allow 
credit be taken for an alternate hydrogen control system in the event 
of both hydrogen recombiners are determined to be inoperable for up to 
7 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with the criteria set forth in 10 CFR 50.92, NMPC 
has evaluated this proposed Technical Specifications change and 
determined it does not represent a significant hazards 
consideration. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change extends the functional test frequency of the 
hydrogen recombiner system. The hydrogen recombiners are not 
considered as initiators for any previously evaluated accidents. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased. The proposed change does not impact the 
Surveillance Requirement itself nor the way in which the 
Surveillance is performed. The proposed change does not affect the 
availability of the hydrogen recombiners to mitigate an accident 
because of the availability of the redundant hydrogen recombiner. 
Furthermore, an historical review of surveillance test results 
indicated that all failures identified were unique, non-repetitive, 
and not related to any time-based failure modes, and indicated no 
evidence of any failures that would invalidate the above 
conclusions. 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 change does not involve any design changes, plant 
modifications, or changes in plant operation. The system will 
continue to function in the same way as before the change. In 
addition, the Surveillance Requirement itself and the way the 
Surveillance is performed will remain unchanged. Furthermore, a 
historical review of surveillance test results indicated no evidence 
of any failures that would invalidate the above conclusions. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The design, function, and OPERABILITY requirements for the 
hydrogen recombiner system are unchanged with this proposed 
revision. Although the proposed change will result in an increase in 
the interval between surveillance tests, the impact on hydrogen 
recombiner availability is small based on the redundant hydrogen 
recombiner, and there is no evidence of any failures that would 
impact the availability of the hydrogen recombiners. Therefore, the 
assumptions in the licensing basis are not impacted, and 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Sheri R. Peterson.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: October 25, 1999.

[[Page 67337]]

    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to add the Oscillation Power 
Range Monitor (OPRM) Upscale function and allow the proposed activation 
of the OPRM function of automatically detecting and suppressing reactor 
instability conditions. Activation of the OPRM is in response to 
Generic Letter 94-02, ``Long-Term Solutions and Upgrade of Interim 
Operating Recommendations for Thermal-Hydraulic Instabilities in 
Boiling Water Reactors,'' licensee's associated commitment to implement 
stability solution Option III as described in Licensing Topical Report 
NEDO-31960-A, ``BWR Owners' Group Long-Term Stability Solutions 
Licensing Methodology,'' and previous Nine Mile Point Unit 2 (NMP2) 
License Amendment 80 dated March 31, 1998. The proposed changes would 
add the OPRM as a Reactor Protection Sytem (RPS) Functional Unit, 
including operability requirements and surveillance tests. 
Specifically, the proposed amendment would revise TS 2.2, ``Limiting 
Safety System Settings,'' TS 3/4.3.1, ``Reactor Protection System 
Instrumentation,'' TS 3/4.4.1, ``Recirculation System,'' and TS 
6.9.1.9, ``Administrative Controls-Core Operating Limits Report.'' The 
proposed changes to support activation of the OPRM function are 
generally consistent with the changes proposed in Licensing Topical 
Report NEDC-32410P-A, ``Nuclear Measurement Analysis and Control Power 
Range Neutron Monitor (NUMAC PRNM) Plus Option III Stability Trip 
Function,'' Supplement 1, dated November 1997. The licensee's submittal 
also provides changes to the associated TS Bases and the TS Index (page 
ix).
    The proposed changes would be made to NMP2's current TS, as well as 
to NMP2's improved TS addressed in a previous notice (64 FR 56518, 
October 20, 1999).
    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 operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The addition of the OPRM Upscale functional unit to TSs involves 
a system that is intended to detect the symptoms of instability 
events and initiate mitigative actions. The worst case failure of 
the system involved would be a failure to initiate mitigative 
actions (i.e., scram), but no failure can cause an accident. The 
removal of certain RCS [Recirculation System] operational 
restrictions is justified with the addition of the OPRM functional 
unit which will provide an automatic scram in the event of reactor 
instabilities. Therefore, the proposed change will not result in a 
significant increase in the probability of any accidents previously 
evaluated.
    The addition of the OPRM Upscale functional unit to the NMP2 TSs 
will permit activation of the OPRM. Activation of the OPRM, together 
with the NUMAC-PRNM, provides NMP2 the ability to detect and 
suppress reactor instabilities. The existing RPS functional units as 
well as other plant equipment will continue to perform their 
intended function in the event of an accident. The addition of the 
OPRM functional unit fulfills the intended purpose of the TS-
required RCS operational restrictions. Therefore, the proposed 
change will not result in a significant increase in the consequences 
of any accident previously evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The addition of the OPRM Upscale functional unit to the NMP2 TSs 
will permit activation of the OPRM. Activation of the OPRM, together 
with the NUMAC-PRNM, provides NMP2 the ability to detect and 
suppress reactor instabilities. The OPRM is a mitigative system 
whose addition as an RPS functional unit will not create the 
possibility of a new or different accident or adversely affect 
existing RPS functional units. The worst case failure of the systems 
involved would be failure to initiate mitigative actions, but no 
failure can cause an accident. Except for the activation of the 
OPRM, no new plant configurations are created. The OPRM Upscale 
functional unit fulfills the intended purpose of the existing TS-
required RCS operational restrictions. Therefore, the proposed 
change will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The proposed TS changes will not adversely affect the 
performance characteristics of RPS instrumentation nor will it 
affect the ability of the subject instrumentation to perform its 
intended function.
    The addition of the OPRM Upscale functional unit to the NMP2 TSs 
will permit activation of the OPRM. Activation of the OPRM, together 
with the NUMAC-PRNM, provides NMP2 the ability to detect and 
suppress reactor instabilities (stability solution Option III) 
thereby meeting the requirements of GDC [General Design Criteria] 10 
and 12. The NRC has reviewed and accepted the Option III methodology 
described in Licensing Topical Report NEDO-31960-A and concluded 
that the solution will provide the intended function. The 
surveillance testing and frequencies proposed will assure 
reliability of the OPRM Upscale function. The purpose of the 
existing TS operational restrictions on the RCS will be met by the 
automatic scram feature of the OPRM.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: October 14, 1999.
    Description of amendment request: The proposed amendments, if 
approved, would revise the LGS, Units 1 and 2, Technical Specifications 
(TSs), Sections 2.2., ``Safety Limits and Limiting Safety System 
Settings,'' and 3.0/4.0, ``Limiting Conditions for Operation and 
Surveillance Requirements.'' The proposed revisions are required to 
support installation of a new Power Range Neutron Monitoring (PRNM) 
System and incorporate long-term thermal-hydraulic stability solution 
hardware.
    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 consequences of an accident previously 
evaluated.
    As discussed in the Nuclear Measurement Analysis & Control 
(NUMAC) PRNM [Power Range Neutron Monitor] Licensing Topical Report 
(LTR), the NUMAC PRNM modification and associated changes to the TS 
involve equipment that is designed to detect the symptoms of certain 
events or accidents and initiate mitigating actions. The worst case 
failure of the equipment involved in the modification is a failure 
to initiate mitigating action (scram or rod block), but no failure 
can cause an accident. The PRNM replacement system is designed to 
perform the same operations as the existing Power Range Monitor 
System and meets or exceeds all operational requirements. Therefore, 
it is concluded that the probability of an accident

[[Page 67338]]

previously evaluated is not increased as a result of replacing the 
existing equipment with the PRNM equipment.
    The PRNM System reduces the need for tedious operator actions 
during normal conditions and allows the operator to focus more on 
overall plant conditions. The automatic self-test and increased 
operator information provided with the replacement system are likely 
to reduce the burden during off-normal conditions as well. The 
replacement equipment qualifications fully envelope the 
environmental conditions, including electromagnetic interference, in 
the LGS control room.
    The replacement equipment has been specifically designed to 
assure that it fully meets the response time requirements in the 
worst case. As a result, due to statistical variations resulting 
from the sampling and update cycles, the response time is typically 
faster than required in order to assure that the required response 
time is always met. Setpoints are changed only when justified by the 
improved equipment performance specifications and by setpoint 
calculations which show that safety margins are maintained. There is 
no impact to the Control Rod Drop accident analysis because the PRNM 
System maintains all existing system functions with a reliability 
equal to or better than the existing Power Range Monitor System.
    The replacement equipment includes up to 5 LPRM [Local Power 
Range Monitor] inputs on a single module compared to one per module 
on the current system. Up to 17 LPRM signals are processed through 
one preprocessor. The recirculation flow signals are processed in 
the same hardware as the LPRM processing. The net effect of these 
architectural aspects is that there are some single failures that 
can cause a greater loss of ``sub-functionality'' than in the 
current system. Other architectural and functional aspects, however, 
have an offsetting effect. Redundant power supplies are used so that 
a single failure of Reactor Protection System (RPS) AC power has no 
effect on the overall PRNM System functions while still resulting in 
a half scram as does the current system. Continuous automatic self-
test also assures that if a single failure does occur, it is much 
more likely to be detected immediately. The net effect is that from 
a total system level, unavailability of the safety-related functions 
in the replacement system is equal to or better than the current 
Power Range Monitor System.
    Based on the extensive and thorough verification and validation 
program used in the PRNM design and field operating experience, 
common cause failures in software controlled functions are judged to 
not be a significant failure mode.
    However, in spite of that conclusion, means are provided within 
the system to mitigate the effects of such a failure and alert the 
operator. Therefore, such a failure, even if it occurred, will not 
increase the consequences of a previously evaluated accident.
    To reduce the likelihood of common cause failure of software 
controlled functions, thorough and careful verification and 
validation activities are performed both for the requirements and 
the implementing software design. In addition, the software is 
designed to limit the loading that external systems or equipment can 
place on the system, thus significantly reducing the risk that some 
abnormal dynamic condition external to the system can cause system 
functional performance problems due to processing ``overload'' 
(i.e., ``slowing down'' or stopping the processing).
    As a conservatism, however, despite these verification and 
validation activities, common cause failures of software-controlled 
functions due to residual software design faults are assumed to 
occur. Both the software and hardware are designed to manage the 
consequences of such failure (and also cover potential common cause 
hardware failures). Safety outputs are designed to be fail safe by 
requiring dynamic update of output modules or data signals, where 
failure to update the information is detected by simple receiving 
hardware, which, in turn, forces a trip. This aspect covers all but 
rather complex failures where the software or hardware executes a 
portion of the overall logic but fails to process some portion of 
new information (inputs ``freeze'') or some portion of the logic 
(outputs ``freeze'').
    To help reduce the likelihood of complex failures, a watchdog 
timer is used which is updated by a very simple software routine 
that in turn monitors the operational cycle time of all tasks in the 
system. The software design is such that as long as all tasks are 
updated at the design rate, it is likely that software controlled 
functions are executing as intended. Conversely, if any task fails 
to update at the design rate, that is a strong indication of at 
least some unanticipated condition. If such a condition occurs, the 
watchdog timer will not be updated, the computer will be 
automatically restarted, and the system will detect an abnormal 
condition and provide an alarm and trip.
    The information available to the operator is at least the same 
as with the current system and, in many cases, improved. No actions 
are required by the operator to obtain information normally used and 
equivalent to that available with the current equipment. However, 
the replacement system does provide more directly accessible 
information regarding the condition of the equipment, including 
automatic self-test, which can aid the operator in diagnosing 
unusual situations beyond those defined in the licensing basis.
    In summary, the reliability of the new PRNM System and its 
ability to detect and mitigate abnormal flux transients have either 
remained the same or improved over the existing Power Range Monitor 
System. Since these postulated reactivity transients are mitigated 
by the new system as effectively and reliability [reliably] as the 
existing system, the consequences of these transients have not 
changed. Therefore, the proposed TS changes do not involve a 
significant increase in the probability 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.
    LGS Modification P00224 uses digital processing with software 
(firmware) control for the main signal processing part of the 
modification. The remainder of the equipment in the modification 
uses conventional equipment similar to the current system (e.g., 
penetrations, cables, interface panels).
    The digital equipment has ``control'' processing points and 
software-controlled digital processing where as the current system 
has analog and discrete component processing. The result is that the 
specific failures of hardware and potential software common cause 
failures are different from the current system. The effects of 
software common cause failure are mitigated by hardware design and 
system architecture, but are of a ``different type'' of failure than 
those evaluated in the LGS Updated Final Safety Analysis Report 
(UFSAR). Therefore, the replacement system may have a malfunction of 
a different type from those evaluated in the LGS UFSAR[. . .] 
However, when these PRNM failures are evaluated at the system level, 
there are no new effects.
    LGS Modification P00224 involves equipment that is intended to 
detect the symptoms of certain transients and accidents and initiate 
mitigating action. The worst case failure of the equipment involved 
in the modification is a failure to initiate mitigating action 
(scram), but no failure can cause an accident. This is unchanged 
from the current system. Software common cause failures could result 
in the system failing to perform its safety function, but this 
possibility is addressed in Section 1, above. In that case, it might 
fail to initiate action to mitigate the consequences of an accident, 
but would not cause one. No new system level failure modes are 
created with the PRNM System.
    Therefore, LGS Modification P00224 does 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 the margin of safety.
    The PRNM System response time and operator information is either 
maintained or improved over the current Power Range Monitor System.
    The PRNM System has improved channel trip accuracy compared to 
the current system and meets or exceeds system requirements assumed 
in setpoint analysis. The channel response time exceeds the 
requirements. The channel indicated accuracy is improved over the 
current system and meets or exceeds all of the system requirements.
    The PRNM System was developed to detect the presence of thermal-
hydraulic instabilities and automatically initiate the necessary 
corrective actions to suppress the oscillations prior to violating 
the Minimum Critical Power Ratio (MCPR) Safety Limit. The NRC has 
reviewed and approved the PRNM Licensing Topical Report (LTR) 
concluding that the PRNM System will provide the intended 
protection.
    Therefore, LGS Modification P00224 does not result in a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 67339]]

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.

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: September 9, 1996, as supplemented on 
June 6, 1997, and June 7, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 Technical Specifications (TSs) proposes to revise TS 
Section 6 to delete requirements for Plant Operating Review Committee 
review of the fire protection program and implementing procedures.
    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:

    Operation of the Indian Point 3 plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes delete the Plant Operating Review Committee 
(PORC) review of changes to the fire protection program and 
implementing procedures. The changes do not introduce any new modes 
of plant operation, make any physical changes, or alter any 
operational setpoints. Therefore, the changes do not degrade the 
performance of any safety system assumed to function in the accident 
analysis. Consequently, there is no effect on the probability or 
consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated.
    No physical changes to the plant or changes to equipment 
operating procedures are proposed. The changes are administrative 
and will not have any direct effect on equipment important to 
safety. Therefore the changes cannot create the possibility of a new 
or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    Adequacy of the fire protection program and implementing 
procedures is assured by the fire protection license condition, the 
procedure review and approval process implemented by Amendment 159, 
the provisions of 10 CFR 50.59, and inspections and audits performed 
under the cognizance of the SRC [Safety Review Committee]. 
Consequently, deleting PORC's responsibility for review of the fire 
protection program and implementing procedure will not degrade the 
fire protection program. Therefore, the proposed changes do not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: November 8, 1999 (PCN 454).
    Description of amendment requests: The licensee proposed to revise 
Surveillance Requirement (SR) 3.8.1.18 of Technical Specification (TS) 
3.8.1, ``A.C. Sources-Operating.'' Currently, SR 3.8.1.18 reads: Verify 
interval between each sequenced load block is within plus or minus 10% 
of design interval for each emergency and shutdown load programmed time 
interval load sequence. The licensee proposed to revise the SR to read: 
Verify the timing of each sequenced load block is within its timer 
setting plus or minus 10% or plus or minus 2.5 seconds, whichever is 
greater, with the exception of the 5 second load group which is minus 
0.5, plus 2.5 seconds, for each programmed time interval load sequence.
    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 any accident previously evaluated?
    Response: No.
    The proposed change would expand the current surveillance 
acceptance criteria to more accurately reflect the characteristics 
of the installed plant equipment. The diesel generators (DG's) have 
sufficient capacity to maintain adequate voltage and frequency 
during load sequencing with the expanded tolerance. The overall 
Engineered Safety Features (ESF) response times in the Technical 
Specifications and safety analyses are maintained even though the 
timer tolerance is increased. Therefore, the consequences of any 
accident previously evaluated are not increased. The DG load 
sequence timers are not of themselves a credible initiator of any 
accident, so the probability of an accident has not been increased. 
The timers will function acceptably to support the equipment needed 
for accident mitigation, so the consequences of an accident are not 
increased. Therefore, the probability or consequences of any 
accident previously evaluated are not increased.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    This amendment request does not involve any change to plant 
equipment or operation. In the event of a loss of preferred power, 
the ESF electrical loads are automatically connected to the DG's in 
sufficient time to provide for safe reactor shutdown and to mitigate 
the consequences of a Design Basis Accident such as a loss of 
coolant accident. Increasing the timer tolerance will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    This amendment does not change the manner in which safety 
limits, limiting safety settings, or limiting conditions for 
operations are determined. The actual response times have not been 
altered by this amendment. Therefore, operation of equipment will 
not be affected. Accordingly, this 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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: November 12, 1999 (PCN 505).
    Description of amendment requests: The licensee proposed to revise 
Technical Specification (TS) 5.5.2.13, ``Diesel Fuel Oil Testing 
Program.'' Specifically, the following changes are proposed:
    1. The at least once per 92 days test is deleted for water and 
sediment,

[[Page 67340]]

American Petroleum Institute (API) gravity or an absolute specific 
gravity, and kinematic viscosity for the diesel fuel oil in the 
Emergency Diesel Generator fuel oil storage tanks. The requirement to 
test these properties prior to addition of new fuel to the storage tank 
remains unchanged.
    2. A requirement is added to test new fuel oil prior to addition to 
the storage tank to verify that the flash point is within limits.
    3. A requirement is added to test new fuel oil within 31 days of 
delivery for ``other properties for ASTM [American Society for Testing 
and Materials] 2D fuel.''
    4. The acceptance criteria for the properties listed, with the 
exception of the particulate criterion, are replaced with the phrase 
``within limits.'' The statement which requires sampling in accordance 
with ASTM-D4057-81 is deleted. Acceptance criteria and reference to the 
applicable standard for sampling are currently provided in the Bases 
for Surveillance Requirement 3.8.3.3.
    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) 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.
    This change is an administrative change to make Technical 
Specification (TS) 5.5.2.13, ``Diesel Fuel Oil Testing Program,'' 
consistent with the existing Bases for Surveillance Requirement (SR) 
3.8.3.3. The specific changes are:
    1. The at least once per 92 days diesel fuel oil test is deleted 
for water and sediment, American Petroleum Institute (API) gravity 
or an absolute specific gravity, and kinematic viscosity. The 
requirement to test these properties prior to addition of new fuel 
to the storage tank remains unchanged.
    2. A requirement is added to test new fuel oil prior to addition 
to the storage tank to verify that the flash point is within limits.
    3. A requirement is added to test new fuel oil within 31 days of 
delivery for ``other properties for ASTM 2D fuel.''
    4. The acceptance criteria for the properties listed, with the 
exception of the particulate content, are replaced with the phrase 
``within limits.'' Acceptance criteria are currently provided in the 
Bases for Surveillance Requirement 3.8.3.3.
    These changes are all consistent with the existing Bases for SR 
3.8.3.3 and NUREG 1432.
    Therefore, this change does 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 kind of 
accident from any accident previously evaluated?
    Response: No.
    This change is an administrative change to make TS 5.5.2.13, 
``Diesel Fuel Oil Testing Program,'' consistent with the existing 
Bases for Surveillance Requirement 3.8.3.3.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    This change is an administrative change to make TS 5.5.2.13, 
``Diesel Fuel Oil Testing Program,'' consistent with the existing 
Bases for Surveillance Requirement 3.8.3.3.
    Therefore, there will be no significant reduction in a margin of 
safety as a result of this change.

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

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

    Date of amendment request: April 19, 1999, as supplemented by 
letter dated November 1, 1999.
    Description of amendment request: The proposed change would revise 
Surveillance Requirement (SR) 3.3.5.2 and associated Bases to allow the 
loss of voltage and degraded voltage trip setpoints to be treated as 
nominal values in the same manner as the trip setpoints for the Reactor 
Trip System (RTS) and Engineered Safety Feature Actuation System 
(ESFAS) instrumentation. The November 1, 1999, letter removes a note 
proposed in the April 19, 1999, amendment request. This revision does 
not change the scope of the April 19, 1999, application and the initial 
proposed no significant hazards consideration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change affects only the presentation of the 
trip setpoints for loss of voltage and degraded voltage in SR 
3.3.5.2 in the VEGP Units 1 and 2 TS [Technical Specifications]. The 
calibration of the channels whose setpoints are specified in SR 
3.3.5.2 will continue to be performed in a manner consistent with 
the setpoint methodology used to determine the trip setpoints. There 
will be no adverse effect on the ability of those channels to 
perform their safety functions as assumed in the safety analyses. 
Since there will be no adverse effect on the trip setpoints or the 
instrumentation associated with those trip setpoints, there will be 
no increase in the probability of any accident previously evaluated. 
Similarly, since the ability of the instrumentation to perform its 
safety function is not adversely affected, there will be no increase 
in the consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change affects only the presentation of the 
trip setpoint requirements of SR 3.3.5.2. Plant operation will not 
be changed, and the response of safety related equipment as assumed 
in the accident analyses would not be adversely affected. Therefore, 
the proposed change does not involve a new or different kind of 
accident than any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. As described above, the loss of voltage and degraded voltage 
instrumentation will remain capable of performing its safety 
function as assumed in the accident analyses. The treatment of trip 
setpoints as nominal values is consistent with the methodology used 
to establish those setpoints. As such, margin is not affected by the 
proposed change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

[[Page 67341]]

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

    Date of amendment request: September 8, 1999, as supplemented by 
letter dated November 9, 1999. The September 8, 1999, application was 
originally noticed in the Federal Register on November 3, 1999 (64 FR 
59806).
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3/4.8.1, ``A.C. Sources, Operating,'' 
and associated Bases, by relocating the 18-month surveillance to 
subject the standby diesel generator to inspections, in accordance with 
procedures prepared in conjunction with its manufacturer's 
recommendations, to the Technical Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, 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 moves the requirement to perform 
manufacturer's recommended inspections of the Standby Diesel 
Generators from the Technical Specifications to the Technical 
Requirements Manual (TRM). The change does not result in any 
hardware or operating procedure changes. The requirement being 
removed from the Technical Specifications is not the initiator of 
any analyzed event. The TRM is maintained using the provisions of 10 
CFR 50.59. Since any changes will be evaluated per 10 CFR 50.59, no 
significant increase in the probability or consequences of an 
accident previously evaluated will be allowed without prior NRC 
approval. Therefore, the changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change moves the requirement to perform 
manufacturer's recommended inspections of the Standby Diesel 
Generators from the Technical Specifications to the TRM. The change 
does not alter the plant configuration (no new or different type of 
equipment will be installed) or make changes in methods governing 
normal plant operation. The change does not impose different 
requirements. The change does not alter assumptions made in the 
safety analysis and licensing basis. Therefore, the change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change moves the requirement to perform 
manufacturer's recommended inspections of the Standby Diesel 
Generators from the Technical Specifications to the TRM. The change 
does not reduce the margin of safety since the location of details 
has no impact on any safety analysis assumptions. In addition, the 
requirement being transposed from the Technical Specification to the 
TRM is the same as the existing Technical Specification. Also, the 
TRM is maintained using the provisions of 10 CFR 50.59. Since any 
changes will be evaluated per 10 CFR 50.59, no significant reduction 
in a margin of safety will be allowed without prior NRC approval.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment requests: November 5, 1999.
    Description of amendment requests: The proposed license amendments 
would revise Technical Specification (T/S) Surveillance Requirement 
4.5.1.c to require verification that power is removed from each 
emergency core cooling system accumulator isolation valve operator 
instead of verification that each accumulator isolation valve breaker 
is removed from the circuit. In addition, the proposed license 
amendments would revise T/S 3.5.1 to change ``pressurizer pressure'' to 
``reactor coolant system pressure'' in the applicability and action 
statement requirements. The Bases for T/S 3/4.5.1 will also be revised 
to reflect both changes. Additionally, administrative changes are 
proposed to the page format.
    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 of occurrence or consequences of an accident previously 
evaluated?
    The ECCS [emergency core cooling system] accumulators are used 
to mitigate the consequences of an accident after the event has 
occurred and do not initiate any accident previously evaluated. 
Demonstrating how power is removed from the valve operator does not 
initiate an accident. Inadvertently closing the valves cannot 
initiate an accident. Therefore, there is no significant increase in 
the probability of occurrence of an accident previously evaluated.
    The ECCS accumulators will still perform their function of 
injecting borated water into the reactor coolant loops following a 
large break loss-of-coolant accident, as described in Section 14.3.1 
of the Updated Final Safety Analysis Report (UFSAR). A spurious 
closure of an accumulator outlet isolation valve is not a credible 
event. Performing T/S Surveillance Requirement 4.5.1.c provides 
assurance that one of the two actions required for spurious closure 
of the valve is precluded. The proposed change to the surveillance 
continues to provide assurance that power will be removed from each 
accumulator isolation valve operator so that the valves remain open. 
The consequences of accidents previously evaluated remained bounded 
because the accumulators will still function as assumed in the UFSAR 
accident analysis. Therefore, there is no significant increase in 
the consequences of any accident previously evaluated.
    Changing ``pressurizer pressure'' to ``RCS [reactor coolant 
system] pressure'' has no significant effect on the applicability of 
the T/S requirements. RCS pressure and pressurizer pressure 
instrumentation measure a similar parameter in the primary coolant 
system. Since the RCS is a closed-loop fluid system, pressure 
instruments should indicate approximately the same value. There is 
no significant difference between the instrument readings because 
they are corrected for range, height, and accuracy. There is no 
significant change in the margin of pressure between when the 
accumulators are required to be aligned at 1000 psig and the upper 
limit specified in T/S 3.5.1.d of 658 psig.
    The proposed format changes are administrative and have no 
impact on plant operation.
    Therefore, the proposed changes do not increase the probability 
of occurrence or

[[Page 67342]]

consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to T/S 3/4.5.1 and the associated Bases do 
not involve any physical changes to the plant, but do change the way 
the plant is operated by changing the method for ensuring spurious 
closure of the accumulator isolation valve will not occur. The 
proposed change to T/S Surveillance Requirement 4.5.1.c does not 
create any new operator actions. The position of the accumulator 
isolation valve remains open in Modes 1, 2, and 3 with RCS pressure 
greater than 1000 psig, which meets its design safety function. The 
proposed change does not increase the possibility of the accumulator 
valve repositioning. In order for repositioning to happen, the 
operator must close the molded-case circuit breaker coupled with 
either an active single failure or deliberate operator action in the 
control room. The proposed change of verifying that power is removed 
from the accumulator isolation valve provides the same level of 
protection. Two positive actions are required for the accumulator 
isolation valve to reposition.
    The proposed format changes are administrative and have no 
impact on plant operation.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    T/S Surveillance Requirement 4.5.1.c provides requirements that 
ensure that a single action will not cause an inadvertent closure of 
the accumulator isolation valves. The proposed change continues to 
ensure that two positive actions, an operator action to restore the 
breaker and a single failure, are required for valve closure.
    Changing ``pressurizer pressure'' to ``RCS pressure'' does not 
impact operation of the accumulators. The proposed changes do not 
impact the nitrogen cover pressure as stated in T/S 3.5.1.c. The 
accumulators would not be expected to inject borated water until RCS 
pressure lowers to 658 psig (the upper limit specified in T/S 
3.5.1.d). The change does not affect when this would occur after an 
accident. Therefore, changing ``pressurizer pressure'' to ``RCS 
pressure'' has no impact on plant operation.
    The proposed format changes are administrative and have no 
impact on plant operation.
    Therefore, there is no significant reduction in the margin of 
safety.

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

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).

Arizona Public Service Company, et al., Docket No. STN 50-528, Palo 
Verde Nuclear Generating Station, Unit No. 1, Maricopa County, Arizona

    Date of application for amendment: October 8, 1999, as supplemented 
October 29, 1999.
    Brief description of amendment: The amendment revises Surveillance 
Requirement 3.8.4.8 of Technical Specification 3.8.4, to allow the 
licensee to forego the performance of this surveillance until entry 
into MODE 4 coming out of the ninth refueling outage for Unit 1.
    Date of issuance: November 19, 1999.
    Effective date: November 19, 1999.
    Amendment No.: 121.
    Facility Operating License No. NPF-41: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 19, 1999 (64 FR 
56369).
    The October 29, 1999, supplement provided clarifying information 
that was within the scope of the original Federal Register notice and 
did not change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 19, 1999.
    No significant hazards consideration comments received: No

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station, 
Unit 1, LaSalle County, Illinois

    Date of application for amendment: July 7, 1999, as supplemented on 
October 14, 1999.
    Brief description of amendment: The amendment revised Section 2.1 
of the Technical Specifications to reflect a change in the Minimum 
Critical Power Ratio.
    Date of issuance: November 9, 1999.
    Effective date: Immediately, to be implemented prior to the startup 
of Cycle 9.
    Amendment No.: 137.
    Facility Operating License No. NPF-11: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 9, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 29, 1998, as 
supplemented by letters dated November 9, 1998, and June 14, 1999.
    Brief description of amendment: This amendment authorized changes 
to the Beaver Valley Power Station, Unit No. 2 (BVPS-2) Updated Final 
Safety Analysis Report (UFSAR). The amendment authorizes changes to the 
UFSAR to reflect revisions to the radiological dose calculations for 
the locked rotor accident analysis. This revision of the calculation 
was performed in order to incorporate more conservative

[[Page 67343]]

assumptions than those used in the previous analysis for a postulated 
locked rotor event.
    These changes are not the result of hardware changes to the plant 
or any change in operating practices. They reflect revised analysis 
results only and allow revision of the licensing basis to reflect 
conservative assumptions used in the revised analyses.
    The June 14, 1999, letter withdrew a portion of the amendment which 
would have revised the UFSAR description of the small-break loss-of-
coolant accident radiological consequences.
    Date of issuance: November 18, 1999.
    Effective date: As of the date of issuance.
    Amendment No: 103.
    Facility Operating License No. NPF-73. Amendment approved changes 
to the UFSAR.
    Date of initial notice in Federal Register: March 11, 1998 (63 FR 
11919).
    The November 9, 1998, and June 14, 1999, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
beyond the scope of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 18, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 29, 1999, as supplemented 
by letters dated August 6, 1999, October 14, 1999, and October 26, 
1999.
    Brief description of amendment: The proposed change to the Arkansas 
Nuclear One, Unit No. 2 Technical Specifications would allow the 
performance of a special inspection of the steam generator tubes during 
an upcoming mid-cycle outage. This mid-cycle outage is planned for the 
purpose of performing inspections in selected areas of the steam 
generator tube bundle where previous inspections have revealed tube 
degradation. The proposed change would limit the initial inspection 
scope to these identified areas and includes scope expansion criteria 
to address unexpected results.
    Date of issuance: November 5, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 210.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54375).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 6, 1999.
    Brief description of amendment: The amendment incorporates the 
Technical Specification changes necessary for redefining the minimum 
critical power ratio safety limit for Cycle 11 operation with a mixed 
core of Siemens Power Corporation fuel and General Electric fuel.
    Date of issuance: November 17, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No: 140.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46434).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 17, 1999.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio.

    Date of application for amendment: July 26, 1999.
    Brief description of amendment: This amendment--
    (1) Relocates the requirements in TS 3/4.3.3.2, ``Instrumentation--
Incore Detectors,'' TS 3/4.3.3.9, ``Instrumentation--Waste Gas System 
Oxygen Monitor,'' and TS 3/4.4.4.7, ``Reactor Coolant System--
Chemistry,'' to the Davis-Besse Nuclear Power Station (DBNPS) Updated 
Safety Analysis Report (USAR) Technical Requirements Manual (TRM);
    (2) Revises TS 3/4.11.2, ``Radioactive Effluents--Explosive Gas 
Mixture,'' to reflect the relocation of TS 3/4.3.3.9;
    (3) Revises the requirements of TS 3/4.4.6.1, ``Reactor Coolant 
System Leakage--Leakage Detection Systems,'' to require one monitor 
(gaseous or particulate) of the containment atmosphere radioactivity 
monitoring systems to be operable, rather than requiring both systems 
to be operable simultaneously; and
    (4) Revises TS 3/4.3.3.1, ``Radiation Monitoring Instrumentation,'' 
to be consistent with the revision to TS 3/4.4.6.1.
    Date of issuance: November 16, 1999
    Effective date: November 16, 1999.
    Amendment No.: 234.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46436).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 16, 1999
    No significant hazards consideration comments received: No

NASA Aeronautics Space Administration (NASA), Docket No. 50-30, NASA 
Test Reactor, Erie County, Ohio

    Date of application for amendment: March 25, 1999, as supplemented 
on August 10, 1999.
    Brief description of amendment: This amendment changes Lewis 
Research Center (LeRC) to Glenn Research Center (GRC).
    Date of issuance: November 16, 1999.
    Effective Date: November 16, 1999.
    Amendment No: 10.
    Facility License No. TR-3: The amendment changes facility name.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54377).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 16, 1999.
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: November 16, 1998, as 
supplemented June 21, 1999.
    Brief description of amendment: Amendment changes Technical 
Specifications to limit reactor power oscillations during a reactor 
trip and allows operation in the Extended Load Line Limit Analysis 
region of the power/flow operating curve.
    Date of issuance: September 21, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 168.

[[Page 67344]]

    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71968) as corrected January 27, 1999 (64 FR 4148).
    The June 21, 1999, letter provided supporting information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 21, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 29, 1998, as supplemented by 
letters dated March 8 and April 7, 1999.
    Description of amendment request: To revise Facility Operating 
License No. NPF-86 to reflect the transfer of the license, to the 
extent held by Montaup Electric Company, to Little Bay Power 
Corporation.
    Date of issuance: November 19, 1999.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 65.
    Facility Operating License No. NPF-86: Amendment revised the 
License.
    Date of initial notice in Federal Register: December 14, 1998 (63 
FR 68801). The March 8 and April 7, 1999 supplements provided 
clarifying information and did not change the staff's proposed no 
significant hazards determination. The Commission received comments 
which were addressed in the staff's Safety Evaluation dated August 3, 
1999. The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 3, 1999.
    No significant hazards consideration comments received: Yes.

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

    Date of application for amendments: April 19, 1999, as supplemented 
August 25, October 14, and November 3, 1999.
    Brief description of amendments: The amendment deletes most of the 
current Technical Specifications to implement the Permanently Defueled 
Technical Specification. Portions of the April 19, 1999, request 
related to fuel storage pool water level, crane operability, and crane 
travel with a spent fuel cask will be addressed at a later date.
    Date of issuance: November 9, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.

Amendment No.: 106.

    Facility Operating License No. DPR-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35208).
    The August 25, 1999, letter provided clarifying information that 
did not change the scope of the April 19, 1999, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 9, 1999.
    No significant hazards consideration comments received: No

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: May 7, 1998, as supplemented 
January 22, 1999.
    Brief description of amendment: The amendment revises the licensing 
basis to address the addition of the dose from the Refueling Water 
Storage Tank back leakage into the design basis loss-of-coolant 
accident analysis and Chapter 15 of the Final Safety Analysis Report.
    Date of issuance: November 4, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 176.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 1, 1998 (63 FR 
35991). The January 22, 1999, supplement provided clarifying 
information that did not change the staff's initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 4, 1999.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: August 5, 1999.
    Brief description of amendment: The amendment corrects editorial 
errors in the Technical Specifications Sections 3.8.3.2, 4.6.2.1, 
4.8.1.1, and 4.9.12. The amendment also corrects minor editorial and 
reference errors in Bases Sections B 3/4.3.2, B 3/4.4.11, B 3/4.6.1.2, 
and B 3/4.8.4.
    Date of issuance: November 15, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 177.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48858).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 15, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 29, 1998, as 
supplemented by letters dated July 30 and October 12, 1999.
    Brief description of amendments: The amendments revise Technical 
Specifications (TS) 6.9.1.8, ``Core Operating Limits Report,'' of the 
current TSs and TS 5.6 of the improved TSs, to allow the use of NRC 
approved addenda to WCAP-10054-P-A, ``Westinghouse Small Break ECCS 
Evaluation Model Using NOTRUMP Code,'' August 1985, to determine core 
operating limits. The improved TSs were issued in Amendment Nos. 135 
for Diablo Canyon Power Plant, Units 1 and 2 dated May 28, 1999, but 
have not yet been implemented.
    Date of issuance: November 15, 1999.
    Effective date: November 15, 1999, and shall be implemented within 
90 days from the date of issuance.
    Amendment Nos.: Unit 1--136; Unit 2-136.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19562). The July 30 and October 12, 1999, supplemental letters provided 
additional clarifying information and did not change the staff's 
initial no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is

[[Page 67345]]

contained in a Safety Evaluation dated November 15, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 12, 1999, as supplemented 
January 29, March 10, and September 20, 1999.
    Description of amendment request: This amendment revised Technical 
Specifications (TSs) Section 3/4.4.2, ``Safety/Relief Valves,'' and TS 
Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2 to increase the 
allowable as-found main steam safety relief valve (SRV) code safety 
function lift setpoint tolerance from plus or minus 1% to plus or minus 
3%. Also, the required number of operable SRVs in operational 
conditions 1, 2, and 3 will be increased from 11 to 12.
    Date of issuance: November 10, 1999.
    Effective Date: As of date of issuance and shall be implemented 
prior to completion of the spring 2000 refueling outage for Limerick 
Generating Station, Unit 1.
    Amendment No: 137.
    Facility Operating License No. NPF-39. The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9194).
    The January 29, March 10, and September 20, 1999, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 10, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 7, 1999.
    Brief description of amendment: The amendment revised the technical 
specifications (TSs) to reflect the permanent deactivation in the 
closed position of the ``wet'' instrument reference leg isolation valve 
HV-61-102. Specifically, TS Table 3.6.3.1, ``Primary Containment 
Isolation Valve,'' and its associated notations were revised to reflect 
this current plant configuration.
    Date of issuance: November 18, 1999.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 138.
    Facility Operating License No. NPF-39. This amendment revised the 
TSs.

    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54380).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 18, 1999.
    No significant hazards consideration comments received: No.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 
3, York County, Pennsylvania

    Date of application for amendments: December 24, 1998, as 
supplemented May 25 and September 27, 1999.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Table 3.3.8.1-1 related to loss of power 
instrumentation set points and limits of allowable values for the 4 kV 
emergency buses.
    Date of issuance: November 16, 1999.
    Effective date: These license amendments are effective as of their 
date of issuance. Phase 1 applies to Functions 2 and 3 in TS Table 
3.3.8.1-1 and shall be implemented within 30 days of the date of 
issuance of the amendment. Phase 2 applies to Functions 4 and 5 in TS 
Table 3.3.8.1-1 and shall be implemented no later than March 1, 2000. 
Note (a) shall be implemented within 30 days of the date of issuance of 
the amendment and shall be voided upon completion of modification 96-
01511, but no later than March 1, 2000.
    Amendments Nos.: 230 and 235.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications. The May 25 and September 27, 
1999, letters provided clarifying information that did not change the 
initial proposed no significant hazards consideration.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24199).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 16, 1999.
    No significant hazards consideration comments received: No.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: October 14, 1997, as 
supplemented July 23, 1998, December 3, 1998, February 25, 1999, and 
September 29, 1999.
    Brief description of amendment: The amendment revises Technical 
Specifications to permit use of additional spent fuel storage racks.
    Date of issuance: November 10, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 256.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 24, 1998 (63 FR 
45096).
    The July 23, 1998, December 3, 1998, February 25, 1999, and 
September 29, 1999, applications provided supplemental information that 
did not affect the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 10, 1999.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: November 14, 1997, as 
supplemented on August 25, 1999.
    Brief description of amendments: The amendments revise the TSs to 
make administrative and editorial changes to correct errors in the TSs 
that have either existed since initial issuance or were introduced 
during subsequent changes. In addition, surveillance requirements are 
added that should have been incorporated within the TSs when the 
applicable amendment to the TSs was approved by the NRC.
    Date of issuance: November 2, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 225 and 206.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (63 
FR 66141). The August 25, 1999, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.

[[Page 67346]]

    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.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of application for amendments: March 18, 1996, as supplemented 
April 28, 1997, and February 16, 1999.
    Brief description of amendment: The amendment authorizes changes to 
the design-basis accident analysis (postulated cask drop accident) to 
be incorporated into the Defueled Safety Analysis Report (DSAR) and 
revises the Permanently Defueled Technical Specifications to reflect 
the changes to the cask drop analysis.
    Date of issuance: November 12, 1999.
    Effective date: November 12, 1999, with the Technical 
Specifications to be implemented within 30 days. Implementation also 
includes incorporation of the changes into the DSAR at the next update 
of the DSAR in accordance with the schedule in 10 CFR 50.71(e).
    Amendment No.: 127.
    Facility Operating License No. DPR-54: The amendment revised the 
Technical Specifications and the Defueled Safety Analysis Report.

    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46442).
    The April 28, 1997, and February 16, 1999, supplements provided 
additional clarifying information that was within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 12, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 20, 1998 (PCN 485), as 
supplemented August 13, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification 3.3.9 by adding a surveillance requirement for response 
time testing for the control room isolation signal.
    Date of issuance: November 15, 1999.
    Effective date: November 15, 1999, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--160; Unit 3--151.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1999 (64 FR 
55311).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 15, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 7, 1998, as supplemented by letters 
dated May 20, June 16, September 30, October 20, and October 21, 1999.
    Brief description of amendments: The amendments changed the 
Technical Specifications (TSs) to reflect reactor coolant system flow 
differences between the existing Model E and replacement Model 
94 steam generators (SGs) by adding a new flow rate 
requirement to TS 3.2.5, Departure from Nucleate Boiling (DNB) 
Parameters, that is applicable to the Model 94 SGs. Related 
changes to Bases 3/4.2.5, DNB Parameters, were also made. The licensee 
withdrew all changes proposed in the May 7, 1998, application that were 
superseded by the previously approved amendments 115/103 dated 
September 2, 1999.
    Date of issuance: November 8, 1999.
    Effective date: November 8, 1999.
    Amendment Nos.: Unit 1--117; Unit 2--105.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 1, 1998 (63 FR 
35996).
    The May 20, June 16, September 30, October 20, and October 21, 
1999, supplements provided additional clarifying information. The 
September 30, 1999, supplement also provided updated TS pages. This 
information was within the scope of the original application and 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 8, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 31, 1998, as supplemented by 
letters dated April 19, August 18, and October 21, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification 3/4.4.9.3 by revising the cold overpressure mitigation 
curve to accommodate the replacement steam generators and by adding two 
surveillances (for the centrifugal charging pumps and the emergency 
core cooling system accumulators) to ensure the operability of the cold 
overpressure mitigation system.
    Date of issuance: November 9, 1999.
    Effective date: November 9, 1999, to be implemented within 30 days.
    Amendment Nos.: Unit 1--118; Unit 2--106.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48867).
    The October 21, 1999, supplement provided a revised implementation 
date. This information was within the scope of the original application 
and Federal Register notice and did not change the staff's initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 9, 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

[[Page 67347]]

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 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 January 3, 2000, 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

[[Page 67348]]

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 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).

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

    Date of amendment request: October 29, 1999, as supplemented 
November 2, 1999.
    Description of amendment request: The amendment revises the 
Technical Specification administrative controls regarding the 
containment leak rate testing program and the core operating limits 
report. These changes are necessary to reflect changes in the accident 
analyses and core design methodologies for the next operating cycle.
    Date of issuance: November 15, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 188.
    Facility Operating License No. DPR-20: Amendment revises the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration, and requested that any comments on 
the proposed no significant hazards consideration be provided to the 
staff by close of business November 12, 1999. The notice was published 
in the Herald Palladium on November 6-8, 1999. No public comments were 
received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
November 15, 1999.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

    Dated at Rockville, Maryland, this 23rd day of November 1999.

    For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-31037 Filed 11-30-99; 8:45 am]
BILLING CODE 7590-01-P