[Federal Register Volume 65, Number 115 (Wednesday, June 14, 2000)]
[Notices]
[Pages 37420-37437]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-14837]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

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

    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.

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.

    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

[[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.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Section Chief: Stephen Dembek

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]]


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

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 -7 per year. This is lower than 
the value determined to be acceptable, i.e., 1  x  10-6 
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.

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

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).

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

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

[[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.

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

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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).
    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.
    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, FQ, 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.

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

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

    Date of amendment request: 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.

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

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

    Date of amendment request: 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.

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

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

[[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 
Reading Room). If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman

[[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