# Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** December 1, 1999
- **Citation:** 64 FR 67330

## Text

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 a

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A99-31037. Public record. Not legal advice.