Google Groups no longer supports new Usenet posts or subscriptions. Historical content remains viewable.

Dismiss

65FR37420 Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, Part 1/3

0 views

Skip to first unread message

robop...@us.govnews.org

unread,

Jun 14, 2000, 3:00:00 AM6/14/00

Archive-Name: gov/us/fed/nara/fed-register/2000/jun/14/65FR37420/part1
Posting-number: Volume 65, Issue 115, Page 37420, Part 1

Message-ID: <65FR...@us.govnews.org>
MIME-Version: 1.0

[Federal Register: June 14, 2000 (Volume 65, Number 115)]
[Notices]
[Page 37420-37437]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr14jn00-120]

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

NUCLEAR REGULATORY COMMISSION

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 20, 2000, through June 2, 2000. The last
biweekly notice was published on May 31, 2000.

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.

[[Page 37421]]

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 July 14, 2000, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and electronically from
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
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. 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).

[[Page 37422]]

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: April 25, 2000.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS)
3/4.9.5, ``Communications'' to allow movement of a control rod in a
fueled core cell in Operational Condition 5, to be exempt from the
communication requirements of TS Section 3/4.9.5 when the control rod
is moved with its normal drive system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
TS Section 3/4.9.5 requires that direct communications be
maintained between the control room and the refueling platform
personnel during Core Alterations in Operational Condition 5. The
requirement to have direct communications maintained between the
control room and the refueling platform personnel does not have an
effect on any accident previously evaluated or the associated
accident assumptions. Thus, the proposed changes do not
significantly increase the probability of an accident previously
evaluated.
The proposed changes do not adversely effect the integrity of
the reactor coolant system or secondary containment. As such, the
radiological consequences of previously evaluated accidents are not
changed. Therefore, the proposed changes do not increase the
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 proposed changes do not affect the assumed accident
performance of any structure, system, or component previously
evaluated. The proposed changes do not introduce any new modes of
system operation or failure mechanisms.
Thus, these proposed changes do 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?
LaSalle County Station, Units 1 and 2, exercise control rods
during Core Alterations in Operational Condition 5. The required
plant conditions for this control rod movement are specified in TS
Section 3/4.9.3, ``Control Rod Position.'' TS Section 3/4.9.3 allows
the movement of one control rod at a time, in a fueled core cell,
under control of the reactor mode switch Refuel position one-rod-out
interlock. The exercising of control rods under the control of the
reactor mode switch Refuel position one-rod-out interlock is
controlled by operators in the control room and does not occur when
fuel is being moved in the reactor pressure vessel (RPV).
The proposed changes do not affect the margin of safety as the
movement of a control rod will continue to satisfy the requirements
of TS Section 3/4.9.3 and will not occur when fuel is being moved in
the RPV.
Thus, this proposed change does not involve a significant
reduction in a margin of safety.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: April 28, 2000.
Description of amendment request: The proposed amendments would
revise License Condition 2.C.(37) for Unit 1 and License Condition
2.C.(21) for Unit 2, to specify the types of fuel movements that cannot
be performed during refueling unless all control rods are fully
inserted.
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:

Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to LaSalle County Station, Unit 1, License
Condition 2.C.(37) and Unit 2 License Condition 2.C.(21), will
require that control rods be fully inserted during the loading and
shuffling of fuel assemblies during refueling in Operation Condition
5. The requirement to have control rods fully inserted during the
loading or shuffling of fuel assemblies, during a refueling in
Operational Condition 5, does not have an effect on any accident
previously evaluated. The removal of fuel assemblies from the RPV
does not affect the initiators or assumptions of a previously
analyzed accident, including inadvertent criticality. Thus, the
probability of the occurrence of an accident previously evaluated is
not increased.
The proposed changes do not affect the analyzed refueling
accidents, the integrity of the Reactor Coolant System or Secondary
Containment. Thus, the radiological consequences of an accident
previously evaluated are not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability of occurrence or consequences of an
accident previously evaluated.
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes to the Unit 1 and 2 License Conditions do
not affect the assumed accident performance of any structure,
system, or component previously evaluated. The proposed changes do
not introduce any new modes of system operation or failure
mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The shutdown margin required during a refueling [outage] is
specified in Technical Specifications (TS) Section 3/4.1.1,
``Shutdown Margin.'' The required shutdown margin ensures that the
core will be maintained sufficiently subcritical to preclude
inadvertent criticality in the shutdown condition. The single
failure inadvertent criticality concerns, during a refueling, are an
unexpected withdrawal of a control rod and the loading of a fuel
assembly into the wrong core cell location. The analysis of these
single failure inadvertent criticality concerns, for a fully loaded
core, has determined that the most limiting event is the unexpected
withdrawal of the highest worth control rod from a fueled cell.
The proposed changes, to the Units 1 and 2 License Conditions,
will prohibit the loading and shuffling of any fuel assembly within
the RPV unless all control rods are fully inserted during a
refueling in Operational Condition 5. The unloading of a fuel
assembly will be consistent with the fuel assembly and control rod
requirements of TS Sections 3/4.9.10.1, ``Single Control Rod
Removal,'' and 3/4.9.10.2, ``Multiple Control Rod Removal.'' These
TS requirements ensure that the proposed changes to the license
conditions will provide assurance that the current analysis for an
unexpected withdrawal of the highest worth control rod from a
totally fueled core remains bounding during a refueling outage.
Thus, the proposed changes do not involve a significant
reduction in a margin of safety.

[[Page 37423]]

Edison Company, P.O. Box 767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: May 1, 2000.
Description of amendment request: The proposed amendments would
revise Technical Specification 3/4.8.1, ``A. C. Sources--Operating,''
to permit functional testing of the emergency diesel generators to be
performed during power operation.
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:

Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The function of the emergency diesel generators (EDGs) is to
supply emergency power in the event of a loss of offsite power.
Operation of the EDGs is not a precursor to any accident. Therefore,
the proposed change to permit the 24-hour functional test of the
EDGs to be performed during power operation does not increase the
probability of an accident previously evaluated.
The EDG that is being tested will be available to supply
emergency loads within the required time to mitigate an accident. In
addition, the remaining required EDGs will be operable during the
test. Furthermore, with any one EDG inoperable the remaining EDGs
are capable of supporting the safe shutdown of the plant. Therefore,
the consequences of an accident previously evaluated are not
significantly changed.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes to the 24-hour functional surveillance test
will not affect the operation of any safety system or alter its
response to any previously analyzed accident. The EDG will
automatically transfer from the test mode of operation, if
necessary, to supply emergency loads in the required time. This mode
of operation is used for the monthly surveillance of the EDGs.
Therefore, no new plant operating modes are introduced.
In the event the EDG fails the functional test, it will be
declared inoperable and the actions required for an inoperable EDG
will be performed. The remaining required EDGs will be maintained
operable and are capable of feeding the loads necessary for safe
shutdown of the plant. This addresses the concerns raised in the NRC
Information Notice 84-69, ``Operation of Emergency Diesel
Generators,'' regarding the operation of EDG[s] connected in
parallel with offsite power. The Information Notice discusses EDG
configurations that have the potential to lead to a complete loss of
offsite and onsite power to safety buses. In summary, the proposed
changes do not adversely affect the performance or the ability of
the EDGs to perform their intended function.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any previously
evaluated.
Do the changes involve a significant reduction in a margin of
safety?
The proposed changes will not reduce availability of the EDG
being tested to provide emergency power in the event of a loss of
offsite power. If a loss of offsite power with a loss of coolant
accident occurs during the surveillance test, the emergency bus
would de-energize and shed load. The EDG would then transfer from
the test mode to the emergency mode. It would then be available to
automatically supply emergency loads. In addition, the remaining
required EDGs would be maintained operable during the test.
Furthermore, with any one EDG inoperable, the remaining EDGs are
capable of supporting the safe shutdown of the plant. The time
required for the EDG being tested to pick up emergency loads will
not be affected by performing the 24-hour functional test during
power operation.
The proposed changes do not affect the assumptions or
consequences of the analyzed accidents. Therefore, the proposed
changes do not change any assumed safety margins.
Therefore, the proposed changes will not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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
Energy Northwest, Docket No. 50-397, WNP-2, Benton County,
Washington
Date of amendment request: April 13, 2000.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement (SR) 3.3.1.1.10
for Function 8 of Table 3.3.1.1-1 and SR 3.3.4.1.2.a. for reactor
protection system (RPS) and end of cycle (EOC) recirculation pump trip
instrumentation to extend the frequency of these SRs from 18 to 24
months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Actuation of the TTV [turbine throttle valve] position switches
is considered in the Turbine Trip accident analysis in Chapter 15 of
the WNP-2 Final Safety Analysis Report. The valve position switches
are assumed to function normally at greater than 30% reactor power
level to initiate a reactor scram to mitigate pressure increase and
an RPT [recirculation pump trip] to terminate jet pump flow in the
accident analysis. The extension of the Channel Calibration
surveillance interval to 24 months does not impact the normal
function of the switches that is assumed in the accident analysis.
There is no increase in probability or consequences represented by
the proposed amendment.
Therefore, the extension of the surveillance intervals does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Historical maintenance and surveillance data indicate there is
no effect on the performance of the TTV position switches resulting
from an extension of the SR interval from 18 to 24 months. To ensure
reliability, WNP-2 periodically replaces the TTV position switches
according to the manufacturers' recommendation. The surveillance
interval extension does not involve a change in design or a change
of switch function. There is no increase in the probability of
failure expected from the interval extension that could result in a
different kind of accident from any previously evaluated.
Therefore, the operation of WNP-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.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Closure of the TTVs isolates the main turbine as a heat sink
producing reactor pressure and neutron flux transients. Eight TTV
limit switches (two per valve) function to actuate RPS and an EOC
RPT to mitigate these transients and terminate jet pump flow. High
pressure and flux transients also actuate RPS resulting in negative
reactivity insertion should there be a failure of the TTV position
switches. Additionally, historical maintenance and surveillance
records indicate that the TTV position switches will operate within
the necessary range and accuracy with the extension of the SR
interval because no position adjustment has been necessary during
past TTV position switch surveillance activities.

[[Page 37424]]

Therefore, operation of WNP-2 in accordance with the proposed
amendment will not involve a significant reduction in the margin of
safety.

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

Date of amendment request: May 8, 2000.
Description of amendment request: The proposed amendment would
change the River Bend Station, Unit 1 (River Bend or RBS), Technical
Specifications (TSs) to remove the Fuel Building and the fuel building
ventilation system from the requirements associated with the Secondary
Containment boundary during operational Modes 1, 2, and 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) The proposed changes, do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the Technical Specifications involve
removing the Fuel Building and the fuel building ventilation system
from the requirements associated with the Secondary Containment
boundary. The changes result in conservatively assuming that all
annulus bypass leakage following a DBA [design basis accident] LOCA
[loss-of-coolant accident] are directed to the environment for the
duration of the accident. Since the proposed changes only affect
functions that are required subsequent to a LOCA or fuel handling
accident (FHA), the proposed changes have no [a]ffect on the
probability of an accident. The Fuel Building portion of the
Secondary Containment boundary is not an active component that could
affect the proper operation of any other essential safety feature or
component. Removal of the Fuel Building from the Secondary
Containment boundary does not affect any other safety-related
system, component, or structure that would increase the probability
of an accident previously evaluated. The proposed change only has an
impact on the dose consequences of the design basis accident and
does not have any affect on the accident precursors or other
accident mitigating features.
A plant-specific radiological analysis has been performed to
assess the affects of the proposed change in the annulus bypass
leakage release pathway in terms of Control Room and off-site doses
following a postulated design basis LOCA. The calculated doses for
all offsite and onsite evaluation points are within the 10 CFR [Code
of Federal Regulations] Part 100 criteria for offsite doses and
within the General Design Criterion 19 of 10 CFR Part 50 for the
Control Room.
The calculated offsite DBA LOCA doses due to the proposed
changes result in an increase of less than 3 percent due to
releasing all annulus bypass leakage directly to the environment.
The control room doses exhibit the largest percentage increase in
the thyroid dose due to the increase in unfiltered and untreated
iodine released to the environment, the release rate to the
environment, and the changes in the control room atmospheric
diffusion coefficient due to dual air intakes. However, the change
in control room thyroid dose reduces the margin to the regulatory
limit by only 4 percent. The calculated doses for all offsite and
onsite evaluation points are not significantly increased and remain
within the 10 CFR Part 100 criteria for offsite doses and within the
General Design Criterion 19 of 10 CFR Part 50 for control room.
The proposed changes also include relaxation of requirements for
the fuel building and fuel building ventilation system except during
the movement of ``recently'' irradiated fuel. The term ``recently
irradiated'' is defined as ``fuel that has occupied part of a
critical reactor core within the previous 11 days.'' This change is
justified based on the irradiated fuel source term decay period.
River Bend currently evaluates three FHA scenarios, one for the fuel
building and two for containment. The FHA-FB [Fuel Building]
scenario would be impacted by the proposed changes since the
scenario assumed filtration for the duration of the release.
However, the proposed changes are bounding in their entirety by the
FHA dose evaluation prepared in support of Amendment 85, as revised
to support Amendment 110. The current analysis assumes that a FHA
occurs with the containment personnel air locks (PAL) open, thus, no
credit is taken for primary containment after an 11-day source term
decay period. The release rate assumed in that analysis bounds the
Fuel Building's normal ventilation rate by a factor of approximately
3 and easily meets Regulatory Guide 1.25 assumptions. All other data
and assumptions (other than decay time of course) are identical for
the two analyses and thus, the Amendment 85 analysis is valid for
the Fuel Building.
It is therefore concluded that the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) The operation of River Bend Station, in accordance with the
proposed amendment, does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes affect the TS requirements for the fuel
building and fuel building ventilation system. These changes have no
impact on any other safety-related system, component, or structure.
The type of accident and the accident precursors are not affected by
changing the annulus bypass release path. The Fuel Building portion
of the Secondary Containment boundary is not an active component
that could affect the proper operation of any other essential safety
feature or component. Also, the accident mitigating features that
are currently credited in the response to the design basis accident
are unchanged by the proposed change. Changing the release path for
the annulus bypass leakage does not create a new or different kind
of accident from the accidents previously evaluated.
It is therefore concluded that the proposed change does not
create the possibility of a new or different kind of accident from
any previously analyzed.
(3) The operation of River Bend Station, in accordance with the
proposed amendment, does not involve a significant reduction in a
margin of safety.
The fuel building and the associated fuel building ventilation
filtration system are currently credited as part of the secondary
containment function. The modified secondary containment boundary
(excluding the fuel building) will still be capable of performing
its design function of limiting offsite and control room dose to
within regulatory limits. The only accident consequences that are
impacted by the proposed change in the secondary containment
(annulus) bypass leakage path are the dose consequences of the
design basis LOCA. The previous dose analysis is changed by assuming
that all annulus bypass leakage is directly to the environment
instead of being released into the Fuel Building where the release
would be treated by the Fuel Building Ventilation System before
release. A plant-specific radiological analysis has been performed
to assess the affects of the proposed change in the annulus bypass
leakage release pathway in terms of Control Room and off-site doses
following a postulated design basis LOCA. The proposed change
required a revision to the existing LOCA dose analysis since the
annulus bypass leakage release is assumed to be directly to the
environment due to removal of the Fuel Building from the Secondary
Containment boundary. The calculated doses for all offsite and
onsite evaluation points are within the 10 CFR Part 100 criteria for
offsite doses and within the General Design Criterion 19 of 10 CFR
Part 50 for the Control Room.
The proposed changes to the Technical Specification requirements
for the fuel building and the fuel building ventilation system when
handling irradiated fuel in the fuel building are bounded by
currently approved FHA analyses.
Therefore, there is no significant reduction in the margin of
safety associated with postulated design basis events at RBS in
allowing the proposed change to the RBS licensing basis.

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

[[Page 37425]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: July 29, 1999, as supplemented by
letters dated August 8, 1999, August 24, 1999, January 27, 2000, March
29, 2000, May 22, 2000, and May 31, 2000.
Description of amendment request: The proposed amendment request
provides additional information to support a modification to Technical
Specification (TS) 3.8.1.1 and associated Bases by extending the
Emergency Diesel Generator (EDG) allowed outage time (AOT) from 72
hours to 10 days. In the supplement letter dated May 22, 2000, an
alternate source for the onsite power system during the EDG maintenance
outage, by way of a temporary EDG (TEDG) has been added. The
application dated July 29, 1999, did not include the TEDG. This notice
supercedes the biweekly Federal Register notice dated February 9, 2000,
(65 FR 6406) based on the original application dated July 29, 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. Will the 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:
The EDGs are backup alternating current power sources designed
to power essential safety systems in the event of a loss of offsite
power. As such, the EDGs are not accident initiators in any accident
previously evaluated. Therefore, this change does not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed changes to the TS will extend the allowed outage
time (AOT) for a single inoperable emergency diesel generator (EDG)
from the current limit of 72 hours to 10 days with the
implementation of compensatory measures. These compensatory measures
consist of a temporary emergency diesel generator (TEDG) capable of
supplying auxiliary power to required safe shutdown loads on the EDG
train removed from service. In the probabilistic risk assessment
(PRA) event of a loss of offsite power, the failure of the operable
EDG, and the failure of the turbine-driven emergency feedwater pump
to start, the TEDG would be started and ready for load within 25
minutes. In the PRA assumptions to calculate the risk increase to
core damage, 50 minutes is available until core uncovery. The AOT
would be extended for: (1) preplanned maintenance work (both
preventive and corrective) known to require greater than 72 hours;
and (2) unplanned corrective maintenance work which may be
determined to take greater than 72 hours.
The plant defense-in-depth has been preserved by the use of a
TEDG to supply required safe shutdown loads. The design basis for
the onsite power systems will continue to conform to 10 CFR 50,
Appendix A, General Design Criterion 17.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will the 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:
The EDGs are backup alternating current power sources designed
to power essential safety systems in the event of a loss of offsite
power. The proposed changes to the TS will extend the allowed outage
time (AOT) for a single inoperable emergency diesel generator (EDG)
from the current limit of 72 hours to 10 days with the
implementation of compensatory measures. These compensatory measures
consist of a temporary emergency diesel generator (TEDG) capable of
supplying auxiliary power to required safe shutdown loads on the EDG
train removed from service. In the PRA event of a loss of offsite
power, the failure of the operable EDG, and the failure of the
turbine-driven emergency feedwater pump to start, the TEDG would be
started and ready for load within 25 minutes. In the PRA assumptions
to calculate the risk increase to core damage, 50 minutes is
available until core uncovery. The AOT would be extended for: (1)
preplanned maintenance work (both preventive and corrective) known
to require greater than 72 hours; and (2) unplanned corrective
maintenance work which may be determined to take greater than 72
hours.
The proposed change does not alter the design, configuration,
and method of operation of the plant for safety-related equipment
during the EDG AOT extension period. The plant defense-in-depth has
been preserved by the use of a TEDG to supply power to required safe
shutdown loads.
The change does involve the modification of non-safety permanent
plant equipment. The modification will involve preparing a 4.16kV
[kilo-volt] non-safety bus breaker for connection to the output of
the TEDG. There is no change being made to the parameters within
which the plant is operated, and the setpoints at which the
protective or mitigative actions initiate. The design basis on which
the plant was licensed will not be changed. In the PRA event of a
loss of offsite power, the failure of the operable EDG, and the
failure of the turbine-driven emergency feedwater pump to start, the
TEDG would be started and ready for load within 25 minutes. In the
PRA assumptions to calculate the risk increase to core damage, 50
minutes is available until core uncovery.
Procedures will be developed to implement onsite power system
recovery action in conjunction with the present Emergency Operating
Procedures (EOP) and appropriate Off Normal Procedures in the event
it is necessary to use the alternate AC power source. The developed
procedures support compensatory measures that provide additional
assurance that if a coincident Loss of Offsite Power and failure of
the operable EDG (outside the design basis of the plant) occurred
during a preplanned maintenance (both preventive and corrective) or
unplanned corrective maintenance extended EDG AOT outage,
appropriate guidance would be available to safely shutdown the
plant. There are no alterations to the existing plant procedure that
will decrease assurance that the plant will remain within analyzed
limits. As such, no new failure modes are being introduced that
would involve any potential initiating events that would create any
new or different kind of accident. The proposed change will only
provide the plant some flexibility in the AOT for accomplishing
preplanned maintenance (both preventive and corrective) normally
performed during refueling outages and any potential unplanned
corrective maintenance that may exceed the normal 72-hour AOT during
plant operation in Modes 1, 2, 3, and 4. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, since there will be no permanent hardware
modifications to safety-related equipment nor alterations in the way
in which the plant or equipment is operated during any design basis
event, the proposed change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Will the operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response:
The proposed change does not affect the LCO's [limiting
conditions for operation] or their Bases used in the deterministic
analysis to establish the margin of safety. The margin of safety is
established through equipment design, operating parameters, and the
setpoints at which automatic actions are initiated. There is no
significant impact on the margin of safety. PSA [probabilistic
safety assessment] methods were used to evaluate the proposed
change. The results of these evaluations indicated the risk
contribution from this proposed AOT with compensatory measures
implemented during this extended EDG AOT time period is small and
within the Regulatory Guide 1.177 risk-informed acceptance
guidelines.
Therefore, the change does not significantly impact the margin
of safety, involve a permanent change in safety-related plant
design, or have any affect on the plant protective barriers.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.

[[Page 37426]]

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

Date of amendment request: October 29, 1999.
Description of amendment request: Entergy Operations, Inc.
(licensee) has proposed to revise their Updated Final Safety Analysis
Report (UFSAR) to discuss the probability threshold for when physical
protection of safety-related components from tornado missiles is
required for certain components. The proposed changes involve the use
of Nuclear Regulatory Commission (NRC) approved probability risk
methodology to assess the need for additional tornado missile
protection and demonstrate that the probability of damage due to
tornado missiles striking safety related components is acceptably low.
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, i.e., revising the current UFSAR
descriptions addressing tornado missile barrier protection at
Waterford Steam Electric Station, Unit 3 (Waterford 3) have been
evaluated against these three criteria, and it has been determined
that the changes do not involve a significant hazard because:
(1) The proposed activity does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The associated UFSAR changes reflect use of the Electric Power
Research Institute (EPRI) Topical Report, ``Tornado Missile Risk
Evaluation Methodology, (EPRI NP-2005),'' Volumes 1 and 2. This
methodology has been reviewed, accepted and documented in a NRC
Safety Evaluation dated October 26, 1983. The NRC concluded that:
``the EPRI methodology can be utilized when assessing the need for
positive tornado missile protection for specific safety-related
plant features in accordance with the criteria of SRP [Standard
Review Plan] Section 3.5.1.4.''
The EPRI methodology has been previously applied by other
licensees to resolve tornado missile protection issues.
The results of the tornado missile hazards analysis are such
that the calculated total tornado missile hazard probability for
safety-related SSC's [systems, structures and components] is
approximately 6.0 x 10 <SUP>-7</SUP> per year. This is lower than
the value determined to be acceptable, i.e., 1 x 10<SUP>-6</SUP>
per year by the NRC Staff.
With respect to the probability of occurrence or the
consequences of an accident previously analyzed in the UFSAR, the
probability of a tornado reaching Waterford 3 causing damage to
plant systems, structures and components is a design basis event
considered in the UFSAR. The changes being proposed herein do not
reduce the probability that a tornado will reach the plant. However,
it was determined that there are a limited number of safety-related
components that theoretically could be struck. The probability of
tornado-generated missile strikes on these components were analyzed
using the NRC Staff approved probability methods described above. On
this basis, the proposed change is not considered to constitute a
significant increase in the probability of occurrence or the
consequences of an accident, due to the low probability of a tornado
missile striking these components.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of previously evaluated
accidents.
(2) The proposed activity does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes involve evaluation of whether any physical
protection of safety-related equipment from tornado missiles is
required relative to the probability of such damage without physical
protection. A tornado at Waterford 3 is a design basis event
considered in the UFSAR. This change involves recognition of the
acceptability of performing tornado missile probability calculations
in accordance with established regulatory guidance.
Therefore, the change would not contribute to the possibility
of, or be the initiator for any new or different kind of accident,
or to occur coincident with any of the design basis accidents in the
UFSAR. The low probability threshold established for tornado missile
damage to system components is consistent with these assumptions.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident.
(3) The proposed activity does not involve a significant
reduction on a margin of safety.
The request does not involve a significant reduction in a margin
of safety. The existing licensing basis for Waterford 3 with respect
to the design basis event of a tornado reaching the plant,
generating missiles and directing them toward safety-related systems
and components is to provide positive missile barriers for all
safety-related systems and components. With the change, it will be
recognized that there is an extremely low probability, below an
established acceptance limit, that a limited subset of the
``important'' systems and components could be struck. The change
from ``protecting all safety-related systems and components'' to
``an extremely low probability of occurrence of tornado generated
missile strikes on portions of important systems and components' is
not considered to constitute a significant decrease in the margin of
safety due to that extremely low probability.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Shippingport, Pennsylvania

Date of amendment request: February 21, 2000.
Description of amendment request: The proposed amendment would
revise the Unit 1 Updated Final Safety Analysis Report (UFSAR)
descriptions for bolting material used on some Reactor Coolant System
(RCS) components.
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 use of carbon steel fasteners in a borated system introduces
a new failure mechanism for the fasteners, that of boric acid
wastage. The materials currently specified in the [Beaver Valley
Power Station] BVPS Unit 1 UFSAR are not susceptible to boric acid
wastage. The probability of failure for all systems may be increased
due to the additional failure mode introduced by change from
corrosion resistant material to carbon steel for RCS and reactor
coolant pressure boundary fasteners.
The design requirements of the [American National Standards
Institute] ANSI and [American Society of Mechanical Engineers] ASME
Codes are conservative in nature, in that, the stress allowable for
fastener materials is less than half the yield strength of the
material, thus creating a margin in the design of two or greater on
structural strength. Therefore, the failure or damage of one or more
non-adjacent fasteners can normally be accommodated. Additionally,
the material properties (Yield and Tensile strength) of the
installed (SA540 Grade B

[[Page 37427]]

Class 23 or 24) carbon steel fasteners are higher than that of the
material identified in the UFSAR (SA453 Grade 660). It should also
be noted that the use of either the carbon steel fasteners (those
installed) or the stainless steel fasteners (those identified in the
UFSAR) is acceptable by the design Codes (ANSI and ASME), the
selection of the material for the fasteners is at the discretion of
the designer and is not specified by Code requirements. When
compared to carbon steel fasteners, the corrosion resistance of
Grade 660 material is pertinent only if leakage is actively
occurring.
The boric acid wastage concern is mitigated by the Boric Acid
Corrosion Program which has systematic measures to ensure that boric
acid corrosion will not lead to degradation of the reactor coolant
pressure boundary. This Boric Acid Corrosion Program with its
inspections provides adequate assurances that abnormal leakage will
be identified and corrective actions taken prior to significant
boric acid corrosion degradation of carbon steel pressure boundary
components.
The NRC, in Generic Letter (GL) 88-05, recognized that boric
acid solution leaking from the reactor coolant system can cause
significant corrosion damage to carbon steel materials. In the GL,
the NRC requested that licensees provide assurance that a boric acid
monitoring program has been implemented. This program was to consist
of systematic measures to ensure that boric acid corrosion does not
lead to degradation of the assurance that the reactor coolant
pressure boundary will have an extremely low probability of abnormal
leakage or rupture. The Beaver Valley Power Station response to the
GL provided assurance that a program was in place and committed to
enhancements to the existing program. An NRC follow-up review was
conducted and the Beaver Valley Power Station program was found to
be acceptable and fulfilling the requirements of GL 88-05
(Reference: NRC Inspection Report Nos. 50-334/88-23 and 50-334/88-
25)
Therefore, the proposed changes to BVPS Unit 1 UFSAR Tables 1.8-
1 and 1.8-2 do not significantly increase the probability or
consequences of any accident previously evaluated in the BVPS Unit 1
UFSAR.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
During an evaluation of the fastener material to be used for the
replacement of a degraded fastener, it was discovered that the BVPS
Unit 1 UFSAR Tables 1.8-1 and 1.8-2 identified that corrosion
resistant materials, SA453 Grade 660, were identified as being
installed. The use of carbon steel fasteners in lieu of the SA453
Grade 660 fasteners identified in the UFSAR introduces the potential
failure mechanism of boric acid corrosion. The corrosion damage that
has occurred on MOV-RC-591 and MOV-CH-310 bolting demonstrates that
corrosion damage from unchecked borated water leakage is damaging to
carbon steel fasteners. Additionally, it should be noted that both
of these degraded conditions were identified and repaired prior to
an operational or structural concern through the application of the
Boric Acid Corrosion Program.
In the design condition (non-corroded), the change to carbon
steel fasteners would not affect the design basis accidents
described in the UFSAR. The boric acid wastage concern is mitigated
by the Boric Acid Corrosion Program which has systematic measures to
ensure that boric acid corrosion will not lead to degradation of the
reactor coolant pressure boundary.
In addition to the Boric Acid Corrosion Program, the body to
bonnet configuration for the fasteners identified in Table 1.8-1 and
1.8-2 result in multiple fasteners for each joint. To meet the
requirements of the design Codes (ANSI or ASME) for valves, the
number of fasteners installed is in excess of the number of
fasteners required to perform the structural function of maintaining
the pressure boundary. Additionally, it is highly unlikely that all
the installed fasteners would corrode in such a manner that
catastrophic failure of the body to bonnet joint would result.
Therefore, the multiple installed fasteners result in an installed
backup to the minimum required number of fasteners necessary to
maintain pressure boundary integrity.
Thus, the assumptions and consequences of the loss of pressure
boundary integrity type of accident would be unchanged and would not
introduce a new or different kind of accident as currently evaluated
in the BVPS Unit 1 UFSAR based on the Boric Acid Corrosion Program
preventing any unacceptable boric acid wastage in accordance with GL
88-05.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change in the Unit 1 UFSAR removing criteria
requiring stainless steel fasteners for RCS and reactor coolant
pressure boundary components would not involve a significant
reduction in the margin of safety since current Technical
Specification requirements remain unchanged and current plant
programs (i.e., Boric Acid Corrosion Program inspections) provide
adequate assurance from the likelihood of corroded fasteners causing
an operational issue. NRC reviewed the Beaver Valley Power Station
Boric Acid Corrosion Program and found the program to be acceptable
to fulfill the requirements of GL 88-05 (Reference: NRC Inspection
Report Nos. 50-334/88-23 and 50-334/88-25).

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

Date of amendment request: May 1, 2000.
Description of amendment request: The proposed amendment would
revise the Unit 1 and 2 Technical Specification (TS) 3/4.6.4.2
Surveillance Requirement (SR). The proposed change would allow
performance of the hydrogen recombiner functional test at containment
pressures greater than the currently specified 13 psia. This would be
accomplished by measuring the flow under normal or current test
conditions (e.g., atmospheric pressure) and calculating the expected
system performance under design basis operating conditions. The
surveillance would be revised to verify that the recombiner flow, when
corrected to the post accident design conditions, is greater than or
equal to the required flow. The corresponding design basis temperature
for post accident recombiner operation would be included in the SR
because it is required to correct the test flow to the design basis
operating conditions. In order to support the calculations necessary to
confirm the recombiner blower performance, the proposed change includes
the addition of an equation and associated discussion to the bases. The
equation will correct the measured test flow to a corresponding flow at
the design basis operating pressure and temperature. In addition to the
technical change described above, SR 4.6.4.2.b.3 would be modified by
separating the criteria for the system blower performance and heater
operation into separate parts of the same surveillance to improve the
presentation of the requirements. Format and editorial changes are
included as necessary to facilitate the revision of the TS text to
conform to the current TS page format, and addition of text to the
bases.
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 does not result in any hardware changes to
the hydrogen recombiners. Additionally, the hydrogen recombiners are
not assumed to be accident initiators of any analyzed event. The
proposed change revises the method for performing the hydrogen
recombiner

robop...@us.govnews.org

unread,

Jun 14, 2000, 3:00:00 AM6/14/00

Archive-Name: gov/us/fed/nara/fed-register/2000/jun/14/65FR37420/part3

Posting-number: Volume 65, Issue 115, Page 37420, Part 1

[[Page 37436]]

of the licensee's application and of the Commission's proposed
determination of no significant hazards consideration. The Commission
has provided a reasonable opportunity for the public to comment, using
its best efforts to make available to the public means of communication
for the public to respond quickly, and in the case of telephone
comments, the comments have been recorded or transcribed as appropriate
and the licensee has been informed of the public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room).

The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By July 14, 2000, the

licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,

NW., Washington, DC and electronically from the ADAMS Public Library

component on the NRC Web site, http://www.nrc.gov (the Electronic

examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.

A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman

[[Page 37437]]

Building, 2120 L Street, NW., Washington, DC, by the above date. A copy

of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of the

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

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

Date of amendment request: May 12, 2000, as supplemented by letter
dated May 19, 2000.
Brief description of amendment: The amendment revises TS 3.7.3,
Condition A, to extend the Completion Time for one or more feedwater
isolation valves (FIVs) inoperable from 4 hours to 24 hours if, within
4 hours, the respective feedwater control valves (FCVs) and the FCV
bypass valves in the same flowpath are verified to be capable of
performing the feedwater isolation function. A footnote is added that
indicates that the extension of the Completion Time to 24 hours is only
applicable for repair of the FIV hydraulic system through fuel cycle 8
for Unit 1 and fuel cycle 5 for Unit 2.
Date of issuance: May 25, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 77.
Facility Operating License Nos. NPF-87 and NPF-89: The amendment
revises the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes The NRC published a public notice of the proposed
amendment, issued a proposed finding of no significant hazards
consideration and requested that any comments on the proposed no
significant hazards consideration be provided to the staff by the close
of business on May 24, 2000. The notice was published in the Dallas
Morning News and the Ft. Worth Star Telegram from May 21 through May
23, 2000.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of Texas, and final
no significant hazards consideration determination are contained in a
Safety Evaluation dated May 25, 2000.

Dated at Rockville, Maryland, this 7th day of June 2000.

For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-14837 Filed 6-13-00; 8:45 am]
BILLING CODE 7590-01-P

robop...@us.govnews.org

unread,

Jun 14, 2000, 3:00:00 AM6/14/00

Archive-Name: gov/us/fed/nara/fed-register/2000/jun/14/65FR37420/part2

Posting-number: Volume 65, Issue 115, Page 37420, Part 1

[[Page 37428]]

functional test specified in Technical Specification (TS)
Surveillance Requirement (SR) 4.6.4.2.b.3. The proposed change to SR
4.6.4.2.b.3 does not reduce the effectiveness of the requirement and
continues to verify the capability of the hydrogen recombiners to
perform their design basis function consistent with the assumptions
of the applicable safety analysis. Therefore, the consequences or
probability of accidents previously evaluated in the UFSAR remain
unchanged.
The addition of supporting TS bases text and the format and
editorial changes made to the TS have no impact on plant operation
or safety.
Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident
previously evaluated.

(2) Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change does not affect any accidents
previously evaluated in the UFSAR and continues to provide assurance
that the hydrogen recombiners remain capable of performing their
design function. The proposed change does not introduce any new
failure modes or affect the probability of a malfunction.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) Does the change involve a significant reduction in a margin
of safety?
The margin of safety depends on the maintenance of specific
operating parameters and systems within design requirements and
safety analysis assumptions.
The proposed change does not involve revisions to any safety
limits or safety system settings that would adversely impact plant
safety. In addition, the proposed change does not affect the ability
of the hydrogen recombiners to perform their design function.

The proposed change revises the method for performing the

hydrogen recombiner functional test specified in SR 4.6.4.2.b.3.
However, the proposed change to SR 4.6.4.2.b.3 does not reduce the
effectiveness of the requirement and continues to verify the
capability of the hydrogen recombiners to perform their design basis
function consistent with the assumptions of the applicable safety
analysis.
The addition of supporting TS bases text and the format and
editorial changes made to the TS have no impact on plant operation
or safety.
Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating

Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Acting Section Chief: Marsha Gamberoni.

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

Date of amendment request: April 27, 2000.

Description of amendment request: The proposed amendment would

change the James A. FitzPatrick Nuclear Power Plant Technical
Specifications by changes to the Trip Level Settings for the Residual
Heat Removal (RHR) and Core Spray (CS) Pump Start Timers as well as the
Automatic Depressurization System (ADS) Auto-Blowdown Timer. The
amendment would also extend the Logic System Functional Test
surveillance test intervals for the RHR, CS and ADS systems from 6
months to 24 months.

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 FitzPatrick 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:
Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This proposed change revises the Trip Level Settings for the RHR
and CS pump interlock start timers as well as the ADS auto-blowdown
timers. This proposed change also extends the surveillance interval
for these timers from 6-months to 24-months.
This proposed change impacts the control of systems designed to
mitigate the consequences of a Loss of Coolant Accident (LOCA).
These changes do not impact any of the Reactor Coolant System
parameter variations listed as potential causes of threats to the
fuel and Reactor Coolant Pressure Boundary listed in section 14.4.2
of the FitzPatrick UFSAR [Updated Final Safety Analysis Report]
(Reference 8) [see application dated April 27, 2000]. Therefore,
this proposed change does not increase the probability of an
accident previously evaluated.
The changes to the control of systems designed to mitigate the
consequences of postulated LOCA events are consistent with the
relevant assumptions made in the FitzPatrick LOCA analysis
(Reference 5) [see application dated April 27, 2000]. Therefore, the
results of that analysis are not changed. Therefore, this proposed
change does not increase the consequence of an accident previously
evaluated. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

This proposed change impacts the control of systems designed to
mitigate the consequences of a Loss of Coolant Accident (LOCA).
These changes do not impact any of the Reactor Coolant System
parameter variations listed as potential causes of threats to the
fuel and Reactor Coolant Pressure Boundary listed in section 14.4.2
of the FitzPatrick UFSAR (Reference 8) [see application dated April
27, 2000]. Therefore, this proposed change does not create the

possibility of a new or different kind of accident from any accident
previously evaluated.

Involve a significant reduction in a margin of safety.
The changes to the control of systems designed to mitigate the
consequences of postulated LOCA events are consistent with the
relevant assumptions made in the FitzPatrick LOCA analysis
(Reference 5) [see application dated April 27, 2000]. Therefore the
results of that analysis are not changed. Therefore, this proposed
change does not reduce the margin of safety.

Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Marsha Gamberoni, Acting.

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

Date of amendment request: February 7, 2000.

Description of amendment request: The proposed amendments would

revise Technical Specification 4.7.1.2.b to make the surveillance
requirements for Auxiliary Feedwater Pump testing consistent with that
of NUREG-1431, ``Standard Technical Specifications, Westinghouse
Plants.'' The Bases associated with this Technical Specification would
also be revised to address the proposed changes.

1. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes to the Technical Specification surveillance
requirements for

[[Page 37429]]

the auxiliary feedwater pumps surveillance testing are consistent
with the latest auxiliary feedwater flow hydraulic model and
accident analyses. The revised minimum acceptance criteria will
ensure that pump degradation, which could adversely impact the
accident analyses, will be detected. The pumps will continue to
operate in the same manner as assumed in the analyses to mitigate
the design basis accidents.
Therefore, there will be no significant increase in the

probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.

The proposed changes to the Technical Specification surveillance
requirements for the auxiliary feedwater pumps surveillance testing
are consistent with the latest auxiliary feedwater flow hydraulic
model and accident analyses. The proposed changes to the Technical
Specification surveillance requirements and associated Bases will
not affect the way the pumps are operated during normal plant
operations or how the pumps will operate after an accident.
Therefore, the proposed changes will not create the possibility

of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.

The proposed changes to the Technical Specification surveillance
requirements for the auxiliary feedwater pumps surveillance testing
are consistent with the latest auxiliary feedwater flow hydraulic
model and accident analyses. The proposed changes to the Technical
Specification surveillance requirements eliminate a potential non-
conservative acceptance value and establish appropriate restrictions
to ensure pump operability. The proposed change to the Technical
Specifications Bases better describes the design function of the
auxiliary feedwater system.
Therefore, there will be no significant reduction in the margin
of safety as defined in the Bases for the Technical Specifications
affected by these proposed changes.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application request: May 25, 2000 (ULNRC-04257).
Description of amendment request: The licensee proposed to
eliminate the technical specifications (TSs) on the boron dilution
mitigation system to avoid a spurious swapover event, such as occurred
during the shutdown for Refueling Outage 9, about 2 years ago. This
amendment would delete the limiting condition for operation, the
actions, and the surveillance requirements for TS 3.3.9, ``Boron
Dilution Mitigation System (BDMS),'' in the instrumentation section of
the TSs for Callaway. In addition, the title of TS 3.3.9 would be
removed from the Table of Contents, the Bases for the TSs would be
revised, and a section on the boron dilution analysis would be added to
Chapter 16 of the Callaway Final Safety Analysis Report (FSAR).

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.

Overall protection system performance will remain within the
bounds of the previously performed accident analyses since the
associated hardware changes described in Section X of Appendix A [to
the application dated May 25, 2000] do not affect any protection
systems. The RTS [reactor trip system] and ESFAS [engineered safety
features actuation system] instrumentation will be unaffected. These
protection systems will continue to function in a manner consistent
with the plant design basis. The installation of an alarm on the
[reactor coolant] letdown divert valve, addition of two redundant
high VCT [volume control tank] water level alarms, and elimination
of the automatic BDMS valve swap-over function will be performed in
such a manner that all design, material, and construction standards
that were applicable prior to the change are maintained.
The proposed change will modify the system interface between
CVCS [chemical and volume control system] and the boron recycle
system such that the RCS [reactor coolant system] and CVCS form a
closed system consistent with the reanalysis assumptions. The
letdown divert valve will be placed in the manual ``VCT'' mode [(1)]
prior to entry into MODE 3 from MODE 2 during a plant shutdown and
[(2)] prior to entry into MODE 5 from MODE 6 during a plant startup
such that letdown flow is directed to the VCT, rather than to the
recycle holdup tanks, except under administrative controls for
planned evolutions which require a high degree of operator
involvement and awareness. These administrative controls will
include verification of the boron concentration of the makeup [to
the reactor coolant] prior to repositioning the divert valve and
restoration requirements to return the valve to the manual ``VCT''
mode upon evolution completion.
The proposed change will not affect the probability of any event
initiators. The above modifications are unrelated to the initiating
event for this analysis, a failure in the reactor makeup control
system. The change will revise the method of detecting the event and
rely on operator action for event termination. There will be no
degradation in the performance of or an increase in the number of
challenges imposed on safety-related equipment assumed to function
during an accident situation. There will be no change to normal
plant operating parameters or accident mitigation performance.
Since manual operator actions are being substituted for
automatic actions, this amendment application was reviewed against
the guidance provided in NRC Information Notice 97-78, ``Crediting
of Operator Actions in Place of Automatic Actions and Modifications
of Operator Actions, Including Response Times.'' Appendix A [to the
application] demonstrates that sufficient time is available for
operator action to terminate the inadvertent boron dilution event
prior to criticality. Additionally, as discussed in NSAC-183, ``Risk
of PWR Reactivity Accidents during Shutdown and Refueling,'' gradual
inadvertent boron dilution events are not expected to cause core
damage, even if they are unmitigated, due to their self-limiting
nature.
The proposed change will achieve the same objective as the BDMS,
i.e., the prevention of an inadvertent criticality as a result of an
unintended boron dilution. The proposed change will not alter any
assumptions or change any mitigation actions in the radiological
consequence evaluations in the FSAR. Appendix A [to the application]
demonstrates that sufficient time is available for operator action
to terminate the inadvertent boron dilution event prior to
criticality. With the reactor subcritical, there will be no increase
in radiological consequences.
Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident
previously evaluated.

2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.

There are no changes in the method by which any safety-related
plant system performs its safety function. The changes described in
Section X of Appendix A [to the application] have no impact on any
analyzed event other than inadvertent boron dilution. The physical
modifications to eliminate the automatic BDMS valve swap-over
function and add redundant high VCT water level alarms and a
position alarm on the letdown divert valve will be implemented in
accordance with existing plant design criteria. The BDMS itself has
no impact on any other analyzed event. The portion of the change
deleting the BDMS from the Technical Specifications, and eliminating
the automatic valve swap-over function, has no other impact safety.
The BDMS flux multiplication alarm will be retained as a plant
design feature to provide the plant operators a diverse method for
identifying a potential dilution event. Since the passive

[[Page 37430]]

alarms to be added only provide information and do not initiate
control or protection system actions, the alarms will not adversely
impact other events. The position of the letdown divert valve only
affects the path for letdown flow. The flow path selected for
letdown does not affect any other accident analyses. Thus, the
operational change to make the manual ``VCT'' mode the normal
operating condition in MODES 3 through 5 has no safety impact.
Procedural changes will heighten the operator awareness of potential
dilution events and provide alarm response actions to mitigate
potential dilution events. As such, these changes will enhance the
response to inadvertent boron dilution events, but have no other
safety impact. The FSAR Chapter 16 requirements for reactor coolant
loop operation and high VCT water level alarm operability will also
enhance the plant operators' capability to respond to an inadvertent
boron dilution event. If the Chapter 16 requirements are not met,
isolating the dilution source valves in MODES 3, 4, and 5 has no
impact on any other accident analyses since none of the other
accident analyses take credit for, or are initiated by, the flow
path through these valves.
This change will affect the normal method of plant operation
while in MODES 3 through 5 with regard to the control of letdown
flow. In these MODES, letdown processing via the recycle holdup
tanks will be allowed only under administrative controls for planned
evolutions which require a high degree of operator involvement and
awareness. The annunication of the letdown divert valve not being in
the ``VCT'' position will further highlight system conditions to the
operating staff. No other performance requirements will be affected.
In order to automatically close the VCT isolation valves, the
RWST [refueling water storage tank] isolation valves must be fully
open. This valve interlock feature is designed to ensure a flow path
is maintained to the CCPs [component cooling pumps] during swap-
over. Since the VCT isolation valves can be manually closed prior to
opening the RWST isolation valves, the possibility exists for the
operator to inadvertently isolate flow to the CCPs while attempting
to isolate the dilution source. However, plant operating procedures
provide the operators with sufficient guidance for performing a
manual valve swap-over and the reanalysis demonstrates that
sufficient time is available to perform the required manual actions,
consistent with SRP [NRC NUREG-0800 Standard Review Plan] acceptance
criteria.
Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.

The proposed change uses acceptance criteria consistent with the
[NRC] Standard Review Plan, as discussed in Appendix A [to the
application]. The margin of safety required of the BDMS is
maintained, i.e., inadvertent boron dilution events will be
terminated by timely operator actions prior to a total loss of all
shutdown margin. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protective functions. There will be no impact
on the overpower limit, DNBR [departure from nucleate boiling ratio]
limits, F<INF>Q</INF>, FdeltaH, LOCA PCT [loss-of-coolant accident
peak cladding temperature], peak local power density, or any other
margin of safety. The radiological dose consequences acceptance
criteria listed in the Standard review Plan will continue to be met.
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied.Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Section Chief: Stephen Dembek.

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

Date of amendment request: May 22, 2000.

Description of amendment request: The proposed amendment would

remove the technical specification surveillance requirement for visual
inspection of suppression chamber coating integrity once each refueling
outage.

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 Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an
accident previously evaluated.

The proposed change conforms the TS to current regulations,
credits actions taken under GL 98-04 to address coating delamination
concerns, and eliminates redundant surveillance criteria. Since
reactor operation under the revised Specification is unchanged, no
design or analytical acceptance criteria will be exceeded. As such,
this change does not impact initiators of analyzed events or assumed
mitigation of accident or transient events. The structural and
functional integrity of plant systems is unaffected. Thus, there is
no significant increase in the probability or consequences of
accidents previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.

The proposed change does not affect any parameters or conditions
that could contribute to the initiation of any accident. No new
accident modes are created. No safety-related equipment or safety
functions are altered as a result of these changes. Because it does
not involve any change to the plant or the manner in which it is
operated, the proposed change does not create the possibility of a

new or different kind of accident from any accident previously
evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

The proposed change does not affect design margins or
assumptions used in accident analyses and has no effect on any
initial condition. The capability of safety systems to function and
limiting safety system settings are similarly unaffected as a result
of this change. Thus, the margins of safety required for safety
analyses are maintained.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.

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

Date of amendment request: May 23, 2000.
Description of amendment request: This proposed change relocates
those portions of Technical Specifications (TSs) related to reactor
coolant conductivity and chloride requirements 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. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an
accident previously evaluated.

The proposed change is administrative in nature and does not
involve the modification of any plant equipment or affect basic
plant operation. Conductivity and chloride limits are not assumed to
be an initiator of any

[[Page 37431]]

analyzed event, nor are these limits assumed in the mitigation of
consequences of accidents.
Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident
previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.

The proposed change does not involve any physical alteration of
plant equipment and does not change the method by which any safety-
related system performs its function. As such, no new or different
types of equipment will be installed, and the basic operation of
installed equipment is unchanged. The methods governing plant
operation and testing remain consistent with current safety analysis
assumptions. Therefore, the proposed change will not create the

possibility of a new or different kind of accident from any accident
previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

The proposed change represents the relocation of current
Technical Specification requirements to the Technical Requirements
Manual, based on regulatory guidance and previously approved changes
for other stations. The proposed change is administrative in nature,
does not negate any existing requirement, and does not adversely
affect existing plant safety margins or the reliability of the
equipment assumed to operate in the safety analysis. As such, there
are no changes being made to safety analysis assumptions, safety
limits or safety system settings that would adversely affect plant
safety as a result of the proposed change. Margins of safety are
unaffected by requirements that are retained, but relocated from the
Technical Specifications to the Technical Requirements Manual.
Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.

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

Date of amendment request: May 23, 2000.

Description of amendment request: The proposed amendment would

revise the technical specification surveillance requirements for local
power range monitor calibration.

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 Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an
accident previously evaluated.

The revised surveillance requirement continues to ensure that
the local power range monitor (LPRM) signal is adequately
calibrated. This change will not alter the basic operation of
process variables, structures, systems, or components as described
in the safety analyses, and no new equipment is introduced by the
change in LPRM surveillance interval. Therefore, the probability of
accidents previously evaluated is unchanged.
The consequences of an accident can be affected by the thermal
limits existing at the time of the postulated accident, but LPRM
chamber exposure has no significant effect on the calculated thermal
limits because LPRM accuracy does not significantly deviate with
exposure. For the extended calibration interval, the total nodal
power uncertainty remains less than the uncertainty assumed in the
thermal analysis basis safety limit, maintaining the accuracy of the
thermal limit calculation. Therefore, the thermal limit calculation
is not significantly affected by LPRM calibration frequency, and the
consequences of an accident previously evaluated are unchanged.
These changes do not affect the initiation of any event, nor do
they negatively impact the mitigation of any event. Therefore, the

proposed changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.

The proposed change will not physically alter the plant or its
mode of operation. As such, no new or different types of equipment
will be installed, and the basic operation of installed equipment is
unchanged. The methods governing plant operation and testing are
consistent with current safety analysis assumptions. Therefore, the
proposed changes will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.
There is no impact on equipment design or fundamental operation,
and there are no changes being made to safety limits or safety
system settings that would adversely affect plant safety as a result
of the proposed change. The margin of safety can be affected by the
thermal limits existing prior to an accident; however, uncertainties
associated with LPRM chamber exposure have no significant effect on
the calculated thermal limits. The thermal limit calculation is not
significantly affected because LPRM sensitivity with exposure is
well defined. LPRM accuracy remains within the total nodal power
uncertainty assumed in the thermal analysis basis, thus maintaining
thermal limits and the safety margin.
Since the proposed changes do not affect safety analysis
assumptions or initial conditions, the margins of safety in the
safety analyses are maintained. Therefore, the proposed changes do

not involve a significant reduction in a margin of safety.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.

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.

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

Date of amendment request: May 4, 2000, as supplemented May 9,

2000.
Description of amendment request: The proposed amendment would

remove the individual control building isolation and recirculation
damper numbers from Technical Specification 4.12.1.3 and instead
specify ``required''

[[Page 37432]]

dampers. The requirement to test these dampers remains the same. The
Bases have been modified to indicate that the damper numbers for
control building isolation and recirculation are contained in the
Updated Final Safety Analysis Report.
Date of publication of individual notice in Federal Register: May
22, 2000 (65 FR 32132).
Expiration date of individual notice: June 21, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

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).

Carolina Power & Light Company, et al., Docket No. 50-325,
Brunswick Steam Electric Plant, Unit 1, Brunswick County, North
Carolina

Date of application for amendment: April 14, 2000, as supplemented
April 20, 2000.
Brief description of amendment: The amendment changed Technical
Specification Surveillance Requirement 3.1.3.3 to allow partial
insertion of control rod 26-47 instead of insertion of one complete
notch. This revised acceptance criterion is limited to the current Unit
No. 1 operating cycle, after which the original one-notch requirement
will be re-established.
Date of issuance: May 23, 2000.
Effective date: May 23, 2000.
Amendment No.: 210.
Facility Operating License No. DPR-71: Amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: April 21, 2000 (65 FR
21481). The April 20, 2000, supplemental letter contained clarifying
information only, and did not change the initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 23, 2000.
No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County,
Washington

Date of application for amendment: July 29, 1999, as supplemented
by letter dated January 31, 2000.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.4.9 applicability from Mode 3 with steam dome
pressure less than residual heat removal cut in permissive to Mode 3
with steam dome pressure less than 48 psig. Notes associated with TS
Surveillance Requirements 3.4.9.1 and 3.5.1.2 are changed to reflect
the new 48 psig limit.
Date of issuance: May 23, 2000.
Effective date: May 23, 2000, to be implemented within 30 days from
the date of issuance.
Amendment No.: 164.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46430).
The January 31, 2000, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 23, 2000.
No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County,
Washington

Date of application for amendment: July 29, 1999, as supplemented
by letters dated August 30, 1999, and February 28, 2000.
Brief description of amendment: The amendment deletes item 3.(b) of
Attachment 2 to License Condition 2.C.(16), that required installation
of a neutron flux monitoring system, in the form of excore wide range
monitors, in conformance with Regulatory Guide 1.97, ``Instrumentation
for Light-Water-Cooled Nuclear Power Plants to Assess Plant and
Environs Conditions During and Following an Accident.''
Date of issuance: May 18, 2000.
Effective date: May 18, 2000, to be implemented within 30 days from
the date of issuance.
Amendment No.: 162.
Facility Operating License No. NPF-21: The amendment revised the
Operating License.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56530).
The February 28, 2000, supplemental letter provided additional
clarifying information but did not expand the scope of the application
as originally noticed and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 18, 2000.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County,
Washington

Date of application for amendment: November 18, 1999, as
supplemented by a letter dated February 7, 2000.
Brief description of amendment: The amendment revised Subsection
4.3.1.2.b of Technical Specification 4.3, ``Fuel Storage.'' The change
revised the wording which described the spacing of the fuel in the new
fuel racks.
Date of issuance: May 23, 2000.
Effective date: May 23, 2000, to be implemented within 30 days from
the date of issuance.
Amendment No.: 163.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.

[[Page 37433]]

Date of initial notice in Federal Register: December 29, 1999 (64
FR 73088)
The February 7, 2000, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May, 23, 2000.
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: November 29, 1999.
Brief description of amendment: The amendment relocated the
requirements associated with the high steam generator level trip
functions of the Reactor Protection System from the Technical
Specifications to the Technical Requirements Manual.
Date of issuance: May 18, 2000.
Effective date: As of the date of issuance to be implemented within

30 days from the date of issuance.

Amendment No.: 216.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 9, 2000 (65 FR
6404).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 18, 2000.
No significant hazards consideration comments received: No.

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

Date of amendment request: August 4, 1999, as supplemented by
letter dated May 18, 2000.
Brief description of amendment: The proposed change modifies the
Technical Specifications (TS) to extend allowed outage time (AOT) to
seven days for one inoperable low pressure safety injection (LPSI)
train. Additionally, an AOT of 72 hours is imposed for other conditions
where the equivalent of 100 percent emergency core cooling system
(ECCS) subsystem flow is available. If 100 percent ECCS flow is
unavailable due to two inoperable LPSI trains, an ACTION has been added
to restore at least one LPSI train to OPERABLE status within one hour
or place the plant in HOT STANDBY within six hours, and to exit the
MODE of applicability within the following six hours. In the event the
equivalent of 100 percent ECCS subsystem flow is not available due to
other conditions, TS 3.0.3 is entered. The Limiting Condition for
Operation terminology is being changed for consistency with the ECCS
requirements. Additionally, the associated TS Bases are being changed.

Date of issuance: May 25, 2000.
Effective date: As of the date of issuance and shall be implemented

60 days from the date of issuance.
Amendment No.: 164.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR
4278).
The May 18, 2000, supplement did not expand the scope of the
application as noticed or change the proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 25, 2000.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 9, 1999, as
supplemented by submittals dated March 1, March 13, and May 11, 2000.
Brief description of amendment: This amendment increases the
present 100 percent authorized rated thermal power level of 3579
megawatts thermal to 3758 megawatts thermal. This represents a power
level increase of 5 percent for the Perry Nuclear Power Plant.
Date of issuance: June 1, 2000.

Effective date: As of the date of issuance and shall be implemented

within 90 days.
Amendment No.: 112.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 3, 1999 (64 FR
59802)
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 1, 2000.
No significant hazards consideration comments received: No.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa

Date of application for amendment: February 18, 1999, as
supplemented September 15, 1999, and March 16, 2000.
Brief description of amendment: The amendment revises Duane Arnold
Energy Center (DAEC) Technical Specification (TS) Table 3.3.6.1-1,
``Primary Containment Isolation Instrumentation,'' by deleting the
manual initiation function of the high pressure coolant injection
(HPCI) system and reactor core isolation cooling (RCIC) system
isolation. A related condition as well as corresponding surveillance
requirements and bases are also deleted.
Date of issuance: June 1, 2000.

Effective date: As of the date of issuance and shall be implemented

within 30 days.
Amendment No.: 231.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17026).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 1, 2000.
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: December 16, 1998.
Description of amendment request: The amendment makes several
editorial and administrative changes to the following sections of the
Technical Specifications (TSs), Index Page vi, ``Figures 3.4-2 and 3.4-
3''; Index Page xv, ``6.0 Administrative Controls''; 4.2.4.2b,
``Determination of Quadrant Power Tilt Ratio''; 6.4.1.7b, ``SORC
Responsibilities''; 6.4.2.2d, ``Station Qualified Reviewer Program'';
6.3.1, ``Training''; 6.4.3.9c, ``Records of NSARC''; 6.8.1.6.b.1,
``Core Operating Limits Report''; and 6.8.1.6.b.10, ``Core Operating
Limits Report''. In addition, the following Bases sections have been
revised: Bases 2.2.1, ``Reactor Trip System Instrumentation
Setpoints''; Bases 3/4.2.4, ``Quadrant Power Tilt Ratio''; Bases 3/
4.2.5, ``DNB Parameters''; Bases 3/4.4.8, ``Specific Activity''; and
Bases 3/4.5.1, ``Accumulators''.
Date of issuance: May 22, 2000.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.

[[Page 37434]]

Amendment No.: 70.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6700).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 22, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendment: April 12, 2000.
Brief description of amendment: The amendment corrects a reference
in Technical Specification Section 6.9.1.8b.1, ``Core Operating Limits
Report.''
Date of issuance: May 26, 2000.

Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.

Amendment No.: 246.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 21, 2000 (65 FR
21486).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 26, 2000.
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: May 19, 2000.
Brief description of amendments: The amendments delay
implementation of the improved Technical Specifications to June 30,
2000 from May 31, 2000.
Date of issuance: May 24, 2000.
Effective date: May 24, 2000.
Amendment Nos.: Unit 1--141; Unit 2--141.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised Appendix D of the Operating Licenses.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendments, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated May
24, 2000.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.

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

Date of application for amendments: May 26, 1999.
Brief description of amendments: The amendments remove Technical
Specification (TS) Surveillance Requirement 4.1.3.5.b, control rod
scram accumulators' alarm instrumentation, and relocate it to the
Updated Final Safety Analysis Report and plant procedures; and revise
TS Action Statement 3.1.3.5.a.2.a to allow for an alternate method of
determining whether a control rod drive pump is operating.
Date of issuance: May 22, 2000.
Effective date: The amendments are effective as of the date of
their issuance and shall be implemented within 30 days. In addition,
the licensee shall include the relocated information in the Updated
Final Safety Analysis Report submitted to the NRC, pursuant to 10 CFR
50.71(e), as was described in the licensee's application dated May 26,
1999 and evaluated in the staff's safety evaluation dated May 22, 2000.
Amendment Nos.: 143 and 105.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 22, 2000 (65 FR
15382).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 22, 2000.
No significant hazards consideration comments received: No.

Portland General Electric Company, et al., Docket No. 50-344,
Trojan Nuclear Plant, Columbia County, Oregon

Date of application for amendment: August 27, 1998, as supplemented
by letter dated July 1, 1999.
Brief description of amendment: The amendment revises the
Permanently Defueled Technical Specifications to delete the requirement
for defueled emergency plan procedures. This amendment is contingent
upon the transfer of the nuclear spent fuel from the existing 10 CFR
Part 50 licensed area to the 10 CFR Part 72 independent spent fuel
storage installation area.
Date of issuance: May 10, 2000.
Effective date: May 10, 2000, and shall be implemented within 30
days after the transfer of the last cask of spent nuclear fuel from the
spent fuel pool to the independent spent fuel storage installation is
complete.
Amendment No.: 202.
Facility Operating License No. NPF-1: The amendment changes the
Permanently Defueled Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46441).
The July 1, 1999, supplemental letter provided additional
clarifying information and did not expand the scope of the application
as originally noticed and did not change the staff's original no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 10, 2000.
No significant hazards consideration comments received: No.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of application for amendments: March 14, 2000, as supplemented
March 27, and May 25, 2000.
Brief description of amendments: The amendments extended the
implementation date for Amendment No. 184 to Facility Operating License
NPF-14 and Amendment No. 158 to Facility Operating License NPF-22 from
30 days following startup from the Unit 1 Spring 2000 refueling outage
to no later than November 1, 2001.
Date of issuance: June 2, 2000.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment Nos.: 187 and 161.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the license.
Date of initial notice in Federal Register: April 27, 2000 (65 FR
24718). The May 25, 2000, letter provided clarifying information but
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 2, 2000.
No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-272, Salem
Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

Date of application for amendment: May 3, 2000, as supplemented on
May 19, 2000.
Brief description of amendment: The license amendment modifies the
existing requirement under Technical

[[Page 37435]]

Specification Section 3.1.3.2.1, Action a.1, to determine the position
of Rod 1SB2 from once every 8 hours to within 8 hours following any
movement of the rod until repair of the rod indication system is
completed. This change is applicable for the remainder of the Unit 1
Cycle 14, or until an outage of sufficient duration occurs whereby the
licensee can repair the position indication system.
Date of issuance: May 26, 2000.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 230
Facility Operating License No. DPR-70: This amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (65 FR 30137) May 10, 2000. The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. That notice also
provided for an opportunity to request a hearing by May 24, 2000, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 26, 2000.

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

Date of application for amendment: January 27, 2000.
Brief description of amendment: The amendment revises the spent
fuel pool reactivity limit requirement by removing the value for K
infinity from Specification 5.6.1.1 and replacing it with a figure of
integral fuel burnable absorbers rods versus nominal Uranium-235
enrichment.
Date of issuance: June 1, 2000.
Effective date: June 1, 2000.
Amendment No.: 144.
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65
FR 9011).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 1, 2000.
No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-260, and 50-296, Browns
Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama

Date of application for amendments: March 15, 2000.
Description of amendment request: The amendments revised the
Technical Specifications (TS) to provide a 7-day limiting condition for
operation when two trains of the Containment Air Dilution System are
inoperable.
Date of issuance: May 24, 2000.
Effective date: May 24, 2000.
Amendment Nos.: 265 and 225.
Facility Operating License Nos. DPR-52 and DPR-68. Amendments
revised the TS.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17919).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 24, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: April 29, 1999.
Brief description of amendments: These amendments revise Technical
Specification (TS) Section 3/4.3.3, ``Radiation Monitoring
Instrumentation,'' TS Section 3/4.7.7, ``Control Room Emergency
Ventilation System,'' and the associated bases. Actions are added and
modified regarding inoperable equipment.
Date of issuance: May 31, 2000.
Effective date: May 31, 2000.
Amendment Nos.: 256 and 247.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27325).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 31, 2000.
No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application for amendment: March 6, 2000 (ULNRC-04197).
Brief description of amendment: The amendment revises Limiting
Condition for Operation (LCO) 3.7.1, ``Main Steam Safety Valves
(MSSVs),'' in that the maximum allowable reactor power for a given
number of operable MSSVs per steam generator is reduced in Table 3.7.1-
1, ``Operable Main Steam Safety Valves [MSSVs] versus Maximum Allowable
Power,'' and in Required Action A.1 of the TSs. These changes will
result in decreasing the setpoint values for the power range neutron
flux high channels, which are part of the reactor trip system (RTS)
instrumentation in Table 3.3.1-1, ``Reactor Trip System
Instrumentation,'' and will result in the reactor operating at a lower
power for a given number of operable MSSVs per steam generator. In
addition, two format errors in the actions for LCO 3.7.1 are corrected.
Date of issuance: May 26, 2000.
Effective date: May 26, 2000, to be implemented within 30 days from
the date of issuance.
Amendment No.: 136.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17920).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 26, 2000.
No significant hazards consideration comments received: No.

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

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility

0 new messages