[Federal Register Volume 64, Number 164 (Wednesday, August 25, 1999)]
[Notices]
[Pages 46424-46455]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-21914]


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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 July 31, 1999, through August 13, 1999. The 
last biweekly notice was published on August 11, 1999 (64 FR 43764).

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

[[Page 46425]]

determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By September 24, 1999, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.714 which 
is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. 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: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public

[[Page 46426]]

document room for the particular facility involved.

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

    Date of amendment request: August 2, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 6.2.2.e to require either the Operations 
Manager or an off-shift Operations superintendent to hold a senior 
reactor operator (SRO) license. This revision would delete the option 
which allows the Manager-Operations to have at one time held a Senior 
Reactor Operator License for a similar unit and replaces it with the 
requirement for an off-shift Operations superintendent who holds an SRO 
license to supervise shift work and licensed activities.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The change to Technical Specification 6.2.2.e to require the 
Manager-Operations or an off-shift Operations superintendent to hold 
an SRO license is administrative in nature and does not directly 
affect plant operations. The change does not physically alter the 
facility in any manner and, as such, does not affect the means in 
which any safety-related system performs its intended safety 
function.
    Therefore, there would be no increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    As stated above, the proposed change is administrative in 
nature. There is no physical alteration to any plant system, nor is 
there a change in the method in which any safety related system 
performs its function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed amendment does not reduce the margin of safety as 
defined in the Safety Analysis Report or the bases contained in the 
Technical Specifications. The requirement to have a licensed SRO 
management position responsible for plant operations is maintained 
within the proposed amendment. The proposed amendment is consistent 
with (1) 10 CFR 50.54(l), which requires individuals responsible for 
directing the licensed activities of licensed operators to hold an 
SRO license, (2) Revision 1 of NUREG-1431, ``Standard Technical 
Specifications Westinghouse Plants,'' and Technical Specification 
Traveler Form (TSTF) 65, Revision 1, and (3) the intent of ANSI/ANS-
3.1, ``Standard for Selection and Training of Personnel for Nuclear 
Power Plants,'' (September 1979 Draft).
    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.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: August 4, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 6.9.1.6.2 to incorporate analytical 
methodology references which are used to determine core operating 
limits. The analytical methodologies to be referenced are documented in 
topical reports which have been accepted by the Nuclear Regulatory 
Commission for referencing in licensing applications.
    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 license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes incorporate additional references to 
methodologies used to evaluate core operating limits. These 
methodologies have been approved for use by the NRC. Plant 
structures, systems, and components will not be operated in a 
different manner as a result of these proposed changes and no 
physical modifications to equipment are involved. Adding these 
references to the Core Operating Limits Report section of Technical 
Specifications does not increase the probability or consequences of 
an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes incorporate additional references to 
methodologies used to evaluate core operating limits. These 
methodologies have been approved for use by the NRC. Plant 
structures, systems, and components will not be operated in a 
different manner as a result of these proposed changes and no 
physical modifications to equipment are involved. Adding these 
references to the Core Operating Limits Report section of Technical 
Specifications does not create the possibility of a new or different 
type of accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes incorporate additional references to 
methodologies used to evaluate core operating limits. These 
methodologies have been approved for use by the NRC. Plant 
structures, systems, and components will not be operated in a 
different manner as a result of these proposed changes and no 
physical modifications to equipment are involved. Adding these 
references to the Core Operating Limits Report section of Technical 
Specifications does not involve a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: May 20, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.8.A to identify the specific 
Containment Cooling Service Water (CCSW) equipment required to support 
operation of the Control Room Emergency Ventilation System (CREVS). The 
proposed amendment would also

[[Page 46427]]

revise TS 3/4.5.C.2 to ensure that the suppression pool water level is 
adequate to prevent vortexing in the Low Pressure Coolant Injection and 
Core Spray pump suctions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    The proposed changes to the technical specifications provide 
clarity in the support system relationship and requirements for the 
CCSW system support of the CREVS operation. [Neither] [t]he CCSW 
system nor the CREVS system are assumed to be accident precursors 
for previously evaluated accident[s]. Therefore, the proposed 
changes have no effect on the probability or consequences of 
accidents previously evaluated.
    The proposed change to the allowable suppression chamber level 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
change revises a Technical Specification acceptance value to [a] 
more conservative value and serves to ensure operability of 
equipment important to safety. By ensuring equipment availability, 
the probability or consequences of an accident previously evaluated 
are not increased. In addition, the proposed changes have no impact 
on any initial condition assumptions for accident scenarios. Onsite 
or offsite dose consequences resulting from an event previously 
evaluated are not affected by this proposed amendment request.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposed changes do not create the possibility of a new or 
different kind of accident from that previously evaluated. The 
changes to the CCSW specifications more appropriate[ly] reflect the 
design requirements and clarify the support role of the CCSW system 
as it relates the CREVS. Neither the CCSW system nor the CREVS will 
be operated differently with the proposed change. Therefore new or 
different failure modes will not be created. Therefore, the 
possibility of new and different accidents has not been created with 
the proposed change. The proposed change to the suppression pool 
allowable level restores margin to the Technical Specifications and 
ensures equipment operability. The proposed change is conservative 
with respect to current requirements. The proposed amendment does 
not involve any plant physical changes that would create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in the margin of safety 
because:
    The proposed change to the CCSW technical specification will not 
result in a significant reduction in the margin of safety. The 
proposed change has greater consistency with the current design 
requirements for CSSW support of CREVS operation. Therefore, the 
margin of safety has been not been altered. [Therefore, the margin 
of safety has not been altered. SIC]
    The proposed changes for suppression pool level does not involve 
a significant reduction in a margin of safety. In fact, the proposed 
changes restore margin and ensure equipment operability. Since the 
changes maintain the necessary level of system reliability, they do 
not involve a significant reduction in the margin of safety.
    The proposed amendment for Dresden will not reduce the 
availability of systems required to mitigate accident conditions; 
therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    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: July 14, 1999.
    Description of amendment request: The proposed amendments would 
allow the units to operate at an uprated power level of 3489 MWt, an 
increase of 5 percent rated core thermal power.
    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?

A. Evaluation of the Probability of Previously Evaluated Accidents

    The proposed power uprate imposes only minor increases in plant 
operating conditions. No change is made to the reactor operating 
pressure. Operation at uprated conditions will result in moderate 
flow increases in those systems associated with the turbine cycle in 
that steam flow increases by approximately six (6)% and feed flow 
increases by approximately six (6)%. The increase in flow in the 
carbon steel piping systems was evaluated for the effect on flow 
induced erosion and corrosion rates and it was confirmed that power 
uprate has no significant effect on flow induced erosion or 
corrosion. The affected systems are currently monitored by the Flow 
Accelerated Corrosion (FAC) program that addresses erosion and 
corrosion concerns. Continued monitoring of the systems provides a 
high level of confidence in the integrity of potentially susceptible 
high energy piping systems.
    Plant systems and components have been verified to be capable of 
performing their intended design functions at uprated power 
conditions. Where necessary, some components will be modified prior 
to implementation of uprated power conditions to accommodate the 
revised operating conditions. The review has concluded that 
operation at power uprate conditions will not affect the reliability 
of plant equipment, and that current Technical Specifications (TS) 
surveillance requirements ensure adequate monitoring of system 
operability. Systems continue to be operated in accordance with 
current design requirements under uprated conditions, therefore no 
new components or system interactions were identified that could 
lead to an increase in accident probability. Changes to reactor 
scram setpoints are such that no significant increase in scram 
frequency due to operation at uprated conditions will occur.

B. Evaluation of the Consequences of Previously Evaluated Accidents

    The radiological consequences due to the Loss of Coolant 
Accident (LOCA) were calculated and are found to be below the 
applicable regulatory limits. The results are presented in Table 9-3 
of Attachment E [of the July 14, 1999 submittal].
    The LOCA radiological consequences have not significantly 
increased due to power uprate, and radiological consequences 
continue to meet established regulatory limits.
    The radiological evaluations for other non-LOCA Design Basis 
Accidents (DBAs) were also performed and the dose consequences for 
these events did not significantly increase. These changes are 
outlined in Section 9.2 of Attachment E and they demonstrate that 
LaSalle County Station (LCS), Units 1 and 2 still meets the 
applicable regulatory limits.

Non-DBA Radiological Doses

    All of the other radiological releases discussed in Updated 
Final Safety Analysis Report (UFSAR) are either unchanged because 
they are not power-dependent, or increase approximately in linear 
proportion to the amount of the uprate. The dose consequences for 
all of the non-LOCA radiological release accident events did not 
significantly increase, and are bounded by the ``LOCA Radiological 
Consequences''

[[Page 46428]]

events discussed above and were shown to meet the current dose 
acceptance limits. These events are discussed in Section 9.2 of 
Attachment E.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The configuration, operation and event response of the LCS, 
Units 1 and 2 systems, structures or components are [unchanged] by 
operation at uprated power conditions. Analysis of transient events 
has confirmed that the same transients remain limiting and that no 
transient event results in a new sequence of events that could lead 
to a new accident scenario.
    An increase in power level will not create a new fission product 
release path, or result in a new fission product barrier failure 
mode. The current fission product barriers consisting of the reactor 
fuel rod cladding, the reactor coolant pressure boundary, and the 
containment structure remain in place. Fuel rod cladding integrity 
is ensured by operating within thermal, mechanical, and exposure 
design limits, and was confirmed for a representative core by 
performance of transient and accident analysis. Cycle specific 
analysis will continue to be performed for each fuel reload to 
demonstrate the compliance with the applicable transient analysis 
criteria and to establish the cycle specific Minimum Critical Power 
Ratio (MCPR) safety limit and fuel operating limits. The integrity 
of the reactor coolant pressure boundary was confirmed by evaluation 
of the bounding overpressurization event and ensuring that the 
corresponding pressure remained below the American Society of 
Mechanical Engineers (AMSE) Boiler and Pressure Vessel (B&PV) Code, 
Section III, ``Rules for Construction of Nuclear Power Plant 
Components,'' overpressure protection requirements. Similarly, 
analysis of the primary containment structure has demonstrated under 
worst case design basis accident conditions that the containment 
structure remains below the containment design pressure.
    The effect of operation at uprated conditions on plant equipment 
has been evaluated. No new operating mode, safety-related equipment 
lineup, accident scenario, or equipment failure mode was identified 
as a result of operating at uprated conditions. In addition, 
operation at power uprated conditions does not create any new 
sequence of events or failure modes that lead to a new type of 
accident. Plant modifications required to support implementation of 
power uprated conditions will be made to existing systems rather 
than by adding new systems of a different design, which might 
introduce new failure modes or accident sequences.
    Therefore, the proposed change does 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 power uprate analysis for LCS, Units 1 and 2 assures that 
the power dependent safety margin will be maintained by meeting the 
appropriate regulatory criteria as prescribed by the applicable 
regulations. Similarly, factors of safety specified by application 
of the regulatory required design rules have been maintained, as 
have other acceptance criteria used to judge the acceptability of 
current plant operation.
    No change is required in the basic duel deign to achieve the 
uprated power levels, or to maintain current operating and safety 
margins. No increase in the allowable peak bundle power is requested 
as a result of operation at uprated conditions. The abnormal 
transients have been evaluated for a representative core 
configuration and confirmed that operation at uprated conditions 
does not have an adverse effect on the operating limit MCPR. No 
change to the Safety Limit MCPR results, thus the margin of safety 
as assured by the safety limit MCPR is maintained. The fuel 
operating limits related to heat generation rate would still be met 
at uprated conditions. Cycle specific analysis will continue to be 
performed for each fuel reload to demonstrate the compliance with 
the applicable transient analysis criteria and to establish the 
cycle specific safety limit and fuel operating limits.
    The Emergency Core Cooling System (ECCS)-LOCA performance has 
been evaluated at power uprated conditions using methodologies that 
have been approved by the NRC for 10CFR50.46, ``Acceptance Criteria 
for Emergency Core Cooling Systems for Light-Water Nuclear Power 
Reactors,'' analysis. The current ECCS performance requirements were 
used in the power uprate analysis. The ECCS-LOCA analysis was 
conducted at 102% of the proposed uprated thermal power in 
accordance with regulatory guidance. The necessary analysis for 
operation of General Electric (GE) fuel under uprated conditions and 
the determination that the peak cladding temperature (PCT) remains 
below the 10CFR50.46 limit of 2200 deg.F have been performed. 
However, LCS Unit 2 currently contains a mixed core of GE and 
Siemens Power Corporation (SPC) fuel. LCS obtained [a] TS amendment 
that allows operation with SPC fuel, and approved the use of the SPC 
analytical methodology. The ECCS-LOCA analysis performed to support 
use of the SPC fuel was conducted at a power level that bounds 102% 
of the proposed uprated power level and determined that the PCT, for 
SPC fuel, remains below the 10CFR50.46 limit of 2200 deg.F. The 
analysis for both GE and SPC fuel types demonstrate all 10CFR50.46 
criteria are met. Therefore, there is no reduction in margin with 
respect to maintaining ECCS performance.
    The margin of safety of the reactor coolant pressure boundary is 
maintained under power uprated conditions. The design pressure of 
the RPV and reactor pressure coolant pressure boundary remains at 
1250 psig. The ASME B&PV Code allowable peak pressure is 1375 psig 
(i.e., 110% of design value), which is the acceptance limit for 
pressurization events. The limiting pressurization event is a Main 
Steam Isolation Valve (MSIV) closure with a failure of valve 
position scram and this event results in a calculated peak RPV 
pressure of 1332 psig at the bottom of the RPV. The peak pressure 
remains below the 1375 psig ASME limit. Therefore, there is no 
decrease in margin of safety in the reactor coolant pressure 
boundary.
    The margin of safety of the containment structure is maintained 
under power uprated conditions. The analyses were conducted using a 
newer NRC-reviewed methodology. The pre-uprated cases were run using 
the new methodology and the re-baselined cases were compared to the 
uprated cases. The short-term containment peak pressure analysis re-
baseline result was 39.3 psig compared to the original analysis of 
39.6 psig. At uprated conditions the peak containment drywell 
pressure would be 39.9 psig, and is below the design value of 45 
psig. The long-term containment suppression pool temperature 
analysis re-baseline result was 190 deg.F compared to the original 
analysis result of 200 deg.F. At uprated conditions the analysis 
concluded that in the event of a LOCA, the calculated peak bulk 
suppression pool temperature would be 193 deg.F. This is less than 
the design temperature of the suppression pool of 275 deg.F, and the 
criteria used to ensure adequate Net Positive Suction Head (NPSH) to 
the ECCS pumps which is 212 deg.F. Therefore, power uprate does not 
challenge the structural integrity of the containment structure and 
ECCS NPSH is assured.
    Therefore, operation at power uprated conditions 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692
    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: August 6, 1999
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3/4.6.4, ``Vacuum Relief'' to remove 
specific operability requirements related to position indication for 
the suppression chamber-drywell vacuum breakers. The amendments also 
reformat the action statements for inoperable vacuum breakers, increase 
the surveillance

[[Page 46429]]

interval for verifying that the vacuum breakers are closed, and delete 
the requirement to verify that the manual isolation valves are closed 
for an inoperable and open vacuum breaker.
    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 do not change the hardware configuration of 
the suppression chamber-drywell vacuum breakers, and the vacuum 
breakers are not considered an initiator in any accident scenario. 
The removal of specific indication requirements and the extension of 
the surveillance interval does not impact the ability of the vacuum 
breakers to perform their safety function. The vacuum breakers 
continue to meet their intended design function. The proposed 
changes do not impact the assumed source term for any analyzed 
accident. Therefore, no increases in the probability of an accident 
or consequences will result due to this proposed change.
    Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes do not involve any physical alterations to 
the suppression chamber-drywell vacuum breakers, or cause any 
changes in the method by which the vacuum breakers or the 
containment vacuum relief system performs their associated design 
basis functions. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed changes do not impact the design function assumed 
for the containment vacuum relief system. The proposed changes do 
not require the vacuum breakers to operate in a condition not 
previously assumed in the facility accident analysis. The 
containment vacuum relief system will continue to operate and 
provide the protection assumed in the accident analysis. In order to 
limit bypass, the vacuum breakers are in a normally closed position. 
These vacuum breakers cannot be permanently placed in the open 
position. The proposed decrease in the surveillance frequency 
verifying the closed vacuum breakers will not increase the risk of 
the vacuum breakers being in the open position, since they will only 
open in response to a pressure differential or manual cycling. 
Therefore, the assurance of the operability of the containment 
vacuum breakers would be the same as provided under current 
Technical Specifications. The containment response analysis is 
unchanged, in that the vacuum breakers protect the containment 
structure, the peak containment pressure remains as calculated, and 
the vacuum breakers continue to maintain bypass leakage rates as 
assumed. Therefore this proposed change does not cause a reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.
    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. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: July 16, 1999.
    Description of amendment request: The proposed change to Technical 
Specification Section 3/4.7.D is to eliminate the limit for any one 
main steam line isolation valve (MSIV) leakage of less than or equal to 
11.5 standard cubic feet per hour (scfh), and to replace that with an 
aggregate value of less than or equal to 46 scfh for all four MSIVs.
    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?
    The proposed changes to the Technical Specifications, Appendix 
A, modifies the allowed leakage limit to an aggregate value with no 
change to the total allowed leakage rate. This change does not 
affect either the automatic or manual features that would close the 
MSIVs. There are no physical changes to the plant and plant 
operations remain unchanged. Therefore, this proposed amendment does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The safety function of the MSIVs is to provide a timely steam 
line isolation to mitigate the release of radioactive steam and 
limit reactor inventory loss under certain accident and transient 
conditions. The MSIVs are designed to automatically close whenever 
plant conditions warrant main steam line isolation. Changing the 
leakage limits to include an aggregate value does not affect the 
isolation function. No new equipment will be installed or utilized, 
and no new operating conditions will be initiated as a result of 
this change. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The total allowed leakage rate for all MSIVs remains unchanged 
at 46 scfh. Therefore, there will be no change in the types or 
significant increase in the amounts of any effluents released 
offsite, and, thus, the radiological analyses remain unchanged and 
within the guidelines of 10 CFR 100 and General Design Criteria 19. 
Therefore, these changes do not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    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.

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

    Date of amendment request: July 27, 1999.
    Description of amendment request: The proposed amendments would add 
a surveillance requirement to verify the Keowee out-of-tolerance logic 
trips and blocks closure of the appropriate overhead or underground 
power path breakers. This logic is being added as part of a 
modification to provide voltage and frequency protection for the Keowee 
Hydro Units to protect them from being exposed to out-of-tolerance 
voltage and frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 46430]]


    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    This change does not create any conditions or events, which lead 
to accidents previously, evaluated in the SAR. The Keowee Hydro 
units are used for mitigation of loss of power scenarios. The 
proposed changes do not change the current function of the Keowee 
Hydro Units. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated. The Keowee Hydro units and their role in the Oconee 
emergency power system currently meet the design/licensing basis 
requirements for the system. There is no adverse affect on 
containment integrity and no new release paths are created. The 
proposed changes do not cause any adverse effects to the Keowee 
single failure design or adversely affect the Keowee start time of 
23 seconds. Therefore, the proposed changes do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated:
    The Keowee Hydro units are used for mitigation of loss of power 
scenarios. No accidents new or different than already evaluated in 
the SAR are postulated as a result of the proposed change. No 
setpoints for parameters, which initiate protective or mitigative 
action, are being changed. Therefore, this proposed amendment does 
not create the possibility of any new or different kind of accident.
    3. Involve a significant reduction in a margin of safety:
    The proposed change does not adversely affect any plant safety 
limits, set points, or design parameters. The change also does not 
adversely affect the fuel, fuel cladding, Reactor Coolant System, or 
containment integrity. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    Duke has concluded, based on the above, that there are no 
significant hazards considerations involved in this amendment 
request.

    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.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

Energy Northwest, (formerly known as the Washington Public Power Supply 
System), Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: July 29, 1999.
    Description of amendment request: The proposed amendment would 
change the applicability of Section 3.4.9 of the Technical 
Specifications (TS) from ``Mode 3 with steam drum pressure less than 
the RHR [residual heat removal] cut in permissive'' to ``Mode 3 with 
steam drum pressure less than 48 psig.'' Notes associated with TS 
Surveillance Requirements 3.4.9.1 and 3.5.1.2 would be changed to 
reflect the proposed 48 psig limit.
    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.
    This change involves further restrictions on the use of RHR in 
the shutdown cooling mode of operation during hot shutdown 
conditions. Chapter 15 of the FSAR [Final Safety Analysis Report] 
defines the start of hot shutdown as the point when generated power 
is below one percent rated power. During entry into hot shutdown 
conditions the RHR system will be aligned in the Low Pressure 
Coolant Injection (LPCI) mode of operation. Thus, it will be aligned 
to provide water to the Reactor Pressure Vessel in the event the 
high pressure systems (HPCS and RCIC) are not able to perform this 
function. The change being proposed here has no impact on loss of 
coolant accidents (LOCAs) requiring mitigation using RHR aligned in 
the LPCI mode of operation.
    During the high pressure portion of the hot shutdown condition, 
intersystem (LOCAs) are a concern. The purpose of the RHR SDC 
Isolation Reactor Pressure--High (cut-in permissive) at 135 psig is 
to prevent over-pressurization of portions of the RHR system. This 
protection is not being modified by this change. The instrumentation 
that provides this protection will continue to function as designed. 
This change only impacts the applicability of Technical 
Specification 3.4.9 and when RHR SDC is required to be operable.
    During hot shutdown the reactor is normally cooled down through 
use of the main steam system and the condenser. Other means of 
cooling are also available using the reactor water cleanup system or 
a combination of emergency core cooling system (ECCS) pumps and 
safety relief valves (SRVs). The RHR system aligned in the SDC mode 
is used at the end of this cooling process to reach cold shutdown 
conditions of less than or equal to 200 deg.F. The change being 
proposed results in the RHR SDC being manually initiated at a lower 
pressure and temperature. This change will have no significant 
impact on the capability to cool the reactor.
    FSAR Chapter 15, ``Accident Analysis,'' describes two events 
associated with the RHR system. FSAR section 15.1.6, ``Inadvertent 
Residual Heat Removal Shutdown Cooling Operation,'' describes the 
impact of system operation during startup or cool-down when the 
reactor is near critical. The proposed change involves the point at 
which RHR is started in the SDC mode with the reactor sub-critical 
with control rods inserted. Therefore, there will be no change in 
the probability or consequences of this accident.
    FSAR section 15.2.9, ``Failure of Residual Heat Removal Shutdown 
Cooling,'' describes the failure of the RHR system to function in 
SDC mode. This evaluation assumes a failure of the SDC mode of 
operation but does not disable the remaining modes of RHR operation. 
The alternate shutdown cooling paths involve the use of the SRVs 
[safety relief valves] to establish a cooling path to the 
containment suppression pool. This evaluated accident does not 
result in any fuel failure. The proposed change will not result in 
any fuel failures. The evaluated accident does result in normal 
coolant activity being released to the suppression pool through the 
safety relief valves. The proposed activity will not result in a 
significant change in the release of this coolant activity.
    The proposed change will not cause a significant increase in the 
probability of a loss of SDC accident. This change proposes a delay 
in the use of SDC because of temperature limitations. During this 
time other means of decay heat removal would be used. This will 
result in a decrease in use of RHR in SDC mode and a decrease in the 
probability of failure of the system by restricting operation to be 
within analyzed temperature limits. The proposed change will not 
involve a significant increase in the consequences of the loss of 
shutdown cooling accident. The accident evaluated in the FSAR 
assumes SDC does not operate at any time and alternate means of 
cooling are evaluated. Section 15.2.9.6 states there is no fuel 
failure and release is limited to normal primary coolant activity to 
the suppression pool. The proposed change results in a short delay 
in the use of SDC because of temperature limitations. The accident 
described in FSAR section 15.2.9 bounds this condition and, as a 
result, there will be no increase in accident consequences.
    With multiple means of reactor water makeup and heat removal 
available the restriction in the use of RHR caused by this change 
will not result in 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.
    The proposed change will not cause any new inadvertent shutdown 
cooling startup, loss of water inventory or loss of cooling 
accidents. New or different inadvertent RHR SDC startup accidents 
are not possible because this change is only a further restriction 
on when the system is operated. The LOCA accidents during Mode 3 are

[[Page 46431]]

bounded by the LOCAs defined for Modes 1 and 2. No new primary sytem 
LOCAs can be initiated because of this change. The purpose of the 
RHR cut-in permissive at 135 psig is to prevent overpressurization 
of portions of the RHR system that could cause an intersystem LOCA. 
This change will not result in a new or different kind of 
intersystem LOCA because this is only a further restriction on RHR 
SDC operation. The use of RHR in the SDC mode is restricted to 
operation at a lower pressure and temperature but other systems are 
available to remove the decay heat. No new or different accidents 
are created because of this change.
    The FSAR section 15.2.9 accident, ``Failure of Resident Heat 
Removal Shutdown Cooling,'' is bounding for all other accidents 
which postulate failure of the capability to remove decay heat. No 
additional accidents resulting in the loss of decay heat removal 
capability will be caused by this change.
    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.
    The proposed amendment will increase the reliability of the RHR 
system when operated in shutdown cooling mode by providing assurance 
that the temperature limits of the piping and pipe supports will not 
be exceeded. The ability to protect against an intersystem LOCA is 
unchanged. The ability to remove decay heat from the reactor is not 
changed by this modification as alternate means of heat removal are 
available. Therefore, operation of WNP-2 in accordance with the 
proposed amendment will not involve a reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

Energy Northwest (formerly known as the Washington Public Power Supply 
System), Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: July 29, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 3.3.5.1-1, ``Emergency Core 
Cooling System (ECCS) Instrumentation Items 1.a, 2.a, 4.a and 5.a,'' to 
change the Reactor Vessel Water Level--Low Low Low, Level 1 allowable 
value from the current value of -148 inches to a new value of -142.3 
inches.
    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.
    This change involves the measurement of water level in the 
Reactor Pressure Vessel (RPV) used to initiate the ECCS. The 
accident evaluated for this condition is the spectrum of loss of 
coolant accidents (LOCA) severe enough to decrease the RPV water 
inventory by a significant amount.
    The additional uncertainty introduced because of harsh 
environmental effects could not be accommodated between the existing 
Technical Specification allowable value and the analytical limit. 
This uncertainty results in a requirement that the ECCS be initiated 
at a slightly higher water level than previously calculated. 
Therefore, operation of WNP-2 in accordance with the proposed 
amendment will 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.
    The proposed change will not create a new or different kind of 
accident since it only makes a small change in the RPV water level 
at which the ECCS is initiated. This change is in the conservative 
direction requiring a greater volume of water in the RPV to 
accommodate the uncertainty associated with the harsh environment of 
the water level sensors.
    The level indicating switches are located on instrument racks in 
the Reactor Building. The harsh environment in this building would 
have no impact on the initial trip needed to initiate the ECCS on 
loss of RPV level since conditions in the Reactor Building would be 
benign at the initial stages of the accident. Only if the Level 1 
trip was reset and initiated after a significant period of time 
would the harsh environmental conditions have an impact on the 
accuracy of the level indicating switches. However, increasing the 
water level at which the ECCS is initiated results in a more 
conservative value that adequately includes post-accident harsh 
environment uncertainties and ensures that the associated analytical 
limit is met.
    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.
    The proposed amendment increases the allowable value for water 
level in the RPV. This small increase will result in an increase in 
the margin of safety. A review of the plant settings for the Level 1 
trip indicated that previous settings were within the new allowable 
value.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

Energy Northwest (formerly known as the Washington Public Power Supply 
System), Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: July 29, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement (SR) 3.5.2.2. 
This requirement verifies the adequacy of the water supply in the 
condensate storage tanks (CSTs) which support operation of the high 
pressure core spray (HPCS) system during Modes 4 and 5. Current 
Technical Specification SR 3.5.2.2 requires that CST water level be 
maintained above 13.25 feet in a single tank or above 7.6 feet in each 
tank if the suppression pool level is below its minimum level. It is 
proposed that the CST water level be maintained above 14.8 feet in a 
single tank or above 9.1 feet in each tank.
    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.
    During Modes 4 and 5 HPCS may be required to provide water to 
the reactor vessel if the water level decreases. The revised 
condensate storage tank allowable levels increase the operating 
margins by providing an increased water inventory. The previously 
evaluated accident involving the loss of decay heat cooling 
inventory will not have an increase in probability because the 
inventory of water will be increased with the change being proposed.

[[Page 46432]]

    The consequences of any accident involving the loss of decay 
heat cooling inventory will not change as the consequences are 
unaffected by the increased water inventory.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will 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.
    The proposed change will not create a new or different kind of 
accident since it only increases the amount of water held in reserve 
to support reactor vessel inventory loss. The proposed change does 
not introduce any credible mechanisms for unacceptable radiation 
release nor does it require physical modification to the plant. The 
inventory of water in the CSTs will increase to support any loss of 
water inventory in the reactor vessel during shutdown.
    The proposed change modifies the monitored values for CST level. 
The plant has operated well within the existing allowable values. 
The increased margin provided by the increased level will assure no 
new or different kinds of accidents result from the proposed change.
    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.
    The proposed amendment increases the allowable value for water 
level in the CSTs. This results in an increase in the inventory of 
water available for cooling and inventory control during reactor 
shutdown. This will result in an increase in the margin of safety.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

Energy Northwest (formerly known as the Washington Public Power Supply 
System), Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: July 29, 1999.
    Description of amendment request: The proposed amendment request 
would revise Technical Specification Surveillance Requirement (SR) SR 
3.8.4.6 of Technical Specification 3.8.4, ``DC Sources--Operating,'' 
and SR 3.8.5.1 of Technical Specification 3.8.5, ``DC Sources--
Shutdown.'' The proposed change to SR 3.8.4.6 would prohibit 
surveillance testing of Division 1, 2, and 3 125 and 250 volt DC, 
battery charger capacity during Modes 1, 2, and 3. However, credit 
could be taken for unplanned events that satisfied the surveillance 
requirement. The proposed change to SR 3.8.5.1 would include SR 3.8.4.6 
as one of the surveillance tests that are not required to be performed.
    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.
    The proposed change has no impact on previously analyzed 
accidents or transients, and has no effect on operation, capacity or 
surveillance test details of the DC system battery chargers. The 
change only imposes a mode restriction on performance of specified 
surveillance testing and allows taking credit for unplanned events 
that satisfy the surveillance. Therefore, operation of WNP-2 in 
accordance with the proposed amendment will 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.
    The proposed change has no effect on operation, capacity, or 
surveillance test details of the DC system battery chargers. The 
change only prohibits performing specified battery charger capacity 
surveillance testing from being implemented during Mode 1, 2, or 3 
and allows taking credit for unplanned events that satisfy the 
surveillance. The proposed change to SR 3.8.4.6 of Technical 
Specification 3.8.4 and SR 3.8.5.1 of Technical Specification 3.8.5 
are consistent with the wording previously evaluated and approved by 
the NRC in NUREG-1434 Rev. 1.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change only imposes a mode restriction, prohibiting 
battery charger capacity surveillance testing from being performed 
during Modes 1, 2, and 3, allowing credit to be taken for unplanned 
events that satisfy the surveillance, and allowing such testing to 
be omitted under certain conditions during Modes 4 and 5 and during 
movement of irradiated fuel in secondary containment. Performance of 
this testing would remove a DC electrical power subsystem from 
service and could present a safety risk were an event to occur if 
the testing was performed in Modes 1, 2, and 3, or while DC service 
is required in other operating conditions. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: July 20, 1998, as supplemented June 29, 
1999.
    Description of amendment request: The amendment would incorporate 
the Technical Specification changes necessary for implementation of the 
Boiling Water Reactor Owners' Group Reactor Stability Long-Term 
Solution, Enhanced Option 1-A (E1A). E1A consists of modifications to 
the plant operating procedures and associated plant components that 
provide a means for reliably detecting and avoiding reactor 
instabilities. By letter dated February 25, 1998, the Nuclear 
Regulatory Commission (NRC) staff recognized E1A as a technically 
acceptable implementation of a long-term stability solution satisfying 
the requirements of NRC IE Bulletin 88-07, Supplement 1, and Generic 
Letter 94-02, ``Long Term Solutions and Upgrade of Interim Operating 
Recommendations for Thermal-Hydraulic Instabilities in Boiling Water 
Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 46433]]


    1. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments allow the implementation of the Enhanced 
Option I-A (E1A) long term solution to the neutronic/thermal-
hydraulic instability issue. Current Technical Specification (TS) 
restrictions on power and flow conditions, number of operating 
recirculation loops and operator actions implemented to reduce the 
probability of neutronic/thermal-hydraulic instability are 
eliminated and new stability requirements consistent with NEDO-
32339-A, Supplement 4, Revision 1, are imposed. These requirements 
include restrictions on power and flow conditions and actions 
associated with the modified Average Power Range Monitor (APRM) flow 
biased scram and control rod block functions. Required actions 
include adherence to the boiling boundary limit stability control 
prior to entry and during operation in the region of the power and 
flow operating domain which is potentially susceptible to neutronic/
thermal-hydraulic instability in the absence of the stability 
control. In addition, the proposed amendments require operator 
actions based upon control room indications generated by a new 
Period Based Detection System (PBDS). The PBDS is designed to 
provide alarm indication that conditions consistent with a 
significant degradation in the stability performance of the reactor 
has occurred and the potential for imminent onset of neutronic/
thermal-hydraulic instability may exist. The PBDS also provides 
analog indication of the highest and second highest successive 
period confirmation count of all of the Local Power Range Monitors 
(LPRMs) monitored. This provides the plant operators with continuous 
indication of reactor stability operating conditions.
    The proposed amendments will permit operation in regions of the 
power and flow operating domain postulated to be susceptible to 
neutronic/thermal-hydraulic instability. Operation in these regions 
does not increase the probability of occurrence of initiators and 
precursors of previously analyzed accidents when neutronic/thermal-
hydraulic instability is not possible. The proposed amendments 
permit the implementation of the features of the E1A solution which 
prevent neutronic/thermal-hydraulic instability including preemptive 
reactor scram upon entry into the regions of the power and flow 
operating domain most susceptible to neutronic/thermal-hydraulic 
instability. The E1A solution also requires implementation of 
stability control prior to entry into a region of the power and flow 
operating domain which is potentially susceptible, in the absence of 
stability control, to neutronic/thermal-hydraulic instability. The 
E1A solution prevents neutronic/thermal-hydraulic instability during 
operation in regions of the power and flow operating domain 
previously excluded from operation and therefore does not 
significantly increase the probability of a previously analyzed 
accident.
    Operation in the regions of the power and flow operating domain 
excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a 
result of anticipated operational occurrences. The severity of these 
transients may increase in the absence of operator actions due to 
the potential occurrence of neutronic/thermal-hydraulic instability 
as a result of operation in these regions. The proposed amendments 
will permit the implementation of the E1A long term solution to the 
stability issue. Required features of the E1A solution include 
adherence to a boiling boundary limit stability control prior to 
selection by the operator of APRM flow biased scram and control rod 
block function ``Setup'' setpoints which allow operation in a region 
of the power and flow operating domain potentially susceptible, in 
the absence of the stability control, to neutronic/thermal-hydraulic 
instability. Upon entry, as a result of an anticipated operational 
occurrence, into the region most susceptible to neutronic/thermal-
hydraulic instability, the preemptive reactor scram prevents 
neutronic/thermal-hydraulic instability. Therefore, the consequences 
of an accident do not significantly increase while operating with 
the stability control met.
    After exiting the region requiring the stability control to be 
met, the setpoints can be manually reset to their normal values. 
Stability controls are required to be in place when setpoints are 
``Setup''. As a backup E1A feature, the APRM flow biased setpoints 
automatically reset to their normal values above a pre-determined 
flow condition. This automatic reset to the more conservative 
setpoints ensures that the preemptive reactor scram will prevent 
operation as a result of an anticipated operational occurrence into 
the region most susceptible to neutronic/thermal-hydraulic 
instability should the operator not select the more conservative 
setpoints appropriate for operation following exit from the region 
requiring stability control.
    Other required E1A features, including the PBDS, control rod 
block alarms associated with entry into the region susceptible to 
neutronic/thermal-hydraulic instabilities in the absence of 
stability controls, and required operator actions, including manual 
reactor scram, help ensure prevention of neutronic/thermal-hydraulic 
instabilities. Therefore, the proposed amendments prevent the 
occurrence of neutronic/thermal-hydraulic instability as a 
consequence of an anticipated operational occurrence and do not 
significantly increase the consequences of any previously analyzed 
accident.
    2. This request does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendments replace current restrictions on power 
and flow conditions with alternative restrictions which permit the 
implementation of the E1A long term stability solution. The current 
restrictions on the power and flow conditions and operating 
recirculation loops in the RUN mode do not automatically prevent the 
entry into regions of the power and flow operating domain most 
susceptible to neutronic/thermal-hydraulic instability and therefore 
the possibility of neutronic/thermal-hydraulic instability exists in 
the absence of operator action. The required features of the E1A 
solution implement a preemptive scram upon entry into the region 
most susceptible to neutronic/thermal-hydraulic instability, without 
operator action. The accessible operating domain allowed by the 
proposed amendments is a subset of the power and flow operating 
domain currently allowed. Current initiators and precursors of 
accidents and anticipated operational occurrences [cannot] occur 
with new or different initial conditions as a result of this change. 
Additionally, there are no new event initiators or precursors of 
accidents and anticipated operational occurrences created by this 
change. Therefore, the proposed amendments do not create the 
possibility of a new or different kind of accident from that 
previously evaluated.
    Concurrent with the implementation of the proposed amendments, a 
modified Flow Control Trip Reference (FCTR) card, the E1A FCTR card, 
and a new Period Based Detection System (PBDS) will be installed as 
required by the E1A solution. The function of the E1A FCTR card is 
to aid the operator in the identification of entry into regions of 
the power and flow operating domain potentially susceptible to 
neutronic/thermal-hydraulic instability in the absence of stability 
controls and to initiate a preemptive scram upon entry into the 
regions most susceptible to neutronic/thermal-hydraulic instability. 
This is accomplished by altering the existing values of setpoints of 
the APRM flow biased scram and the control rod block functions 
generated by the E1A FCTR card. The E1A FCTR card design includes 
components which may be susceptible to electromagnetic interference 
or other environmental effects. The plant specific environmental 
conditions (temperature, humidity, pressure, seismic, and 
electromagnetic compatibility) have been confirmed to be enveloped 
by the environmental qualification values for the E1A FCTR cards. 
Therefore, the potential for spurious scrams or common mode failures 
induced by environmental effects (e.g., electromagnetic 
interference) is considered negligible. The installation of the E1A 
FCTR card will therefore not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The function of the PBDS is to provide the operator with an 
indication that conditions consistent with a significant degradation 
in the stability performance of the reactor has occurred and the 
potential for imminent onset of neutronic/thermal-hydraulic 
instability may exist. This is accomplished by the installation of a 
new PBDS card in the Neutron Monitoring System. The PBDS card takes 
inputs from individual local power range monitors and provides 
analog indication of the highest and second highest successive 
period confirmation count, provides a High Decay Ratio (Hi DR) and 
High-High Decay Ratio (Hi-Hi DR) alarms, and INOP status indication 
to the operator in the control room. These displays [cannot] create 
the possibility of a new or different kind of accident from any 
accident previously evaluated. The PBDS card design includes 
components which may be susceptible to electromagnetic interference 
or other environmental effects. However, the plant specific 
environmental conditions (temperature, humidity, pressure, seismic,

[[Page 46434]]

and electromagnetic compatibility) have been confirmed to be 
enveloped by the PBDS environmental qualification values. Therefore, 
the installation of the PBDS card will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. This request does not involve a significant reduction in a 
margin to safety.
    The proposed amendments permit the implementation of the E1A 
long term solution to the stability issue. Under certain conditions, 
existing BWR [boiling water reactor] designs are susceptible to 
neutronic/thermal-hydraulic instability. General Design Criterion 
(GDC) 12 of 10 CFR 50, Appendix A, requires thermal-hydraulic 
instability to be prevented by design or be readily and reliably 
detected and suppressed. When the design of the reactor system does 
not prevent the occurrence of neutronic/thermal-hydraulic 
instability, instability is an anticipated operational occurrence. 
GDC 10 of 10 CFR 50, Appendix A, requires that specified acceptable 
fuel design limits not be exceeded during anticipated operational 
occurrences.
    Analyses performed by the BWROG [Boiling Water Reactor Owners' 
Group] indicate that neutronic/thermal-hydraulic instability induced 
power oscillations could result in conditions exceeding the Minimum 
Critical Power Ratio (MCPR) Safety Limit (SL) prior to detection and 
suppression by the current design of the Neutron Monitoring System 
and Reactor Protection System.
    To ensure compliance with GDC 12 the BWROG developed Interim 
Corrective Actions (ICAs) to enhance the capability of the operator 
to readily and reliably detect and suppress neutronic/thermal-
hydraulic instability. The BWROG ICAs also provided additional 
guidance for monitoring local power range monitors beyond the 
requirements of current TS 3.4.1 to ensure adequate margin to the 
onset of neutronic/thermal-hydraulic instability. Reliance on 
operator actions to comply with GDC 12 was accepted on an interim 
basis by the NRC pending final implementation of a long term 
solution to the stability issue. Neutronic/thermal-hydraulic 
instability is prevented by implementation of the E1A solution 
through the modified design of the Reactor Protection System (APRM 
[average power range monitor] flow biased scram) and the stability 
control prior to entry into a region of the power and flow operating 
domain which is potentially susceptible, in the absence of stability 
control, to neutronic/thermal-hydraulic instability. In addition, 
significant backup protection features, including the PBDS, control 
rod block alarms associated with entry into the region susceptible 
to neutronic/thermal-hydraulic instabilities in the absence of 
stability controls, and specified operator actions, including manual 
reactor scram, are required to be implemented. As a result, the 
margin to the onset of neutronic/thermal-hydraulic instability 
provided by the existing TS requirements and BWROG ICAs 
recommendations is not significantly reduced by the implementation 
of the E1A solution. The E1A solution assures compliance with GDC 12 
by the prevention of neutronic/thermal-hydraulic instability and 
therefore precludes neutronic/thermal-hydraulic instability from 
becoming a credible consequence of an anticipated operational 
occurrence. The consequences of anticipated operational occurrences 
will not increase and the margin to the MCPR SL will not decrease 
upon implementation of the E1A solution. Therefore, the proposed 
amendments 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.
    Local Public Document Room Location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 6, 1999.
    Description of amendment request: The proposed amendments would 
change those Technical Specifications (TS) required to support Grand 
Gulf Nuclear Station (GGNS), Cycle 11 operation. The changes would 
include a change to the minimum critical power ratio safety limit 
(SLMCPR) that would reflect a decrease of the two recirculation loop 
SLMCPR limit from 1.11 to 1.09, and the single recirculation loop 
SLMCPR limit from 1.12 to 1.10. These values were developed with 
General Electric's cycle-specific SLMCPR methodology in GESTAR-II 
Amendment 25, which was recently approved by the Nuclear Regulatory 
Commission in a Safety Evaluation Report dated March 11, 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:

    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The Minimum Critical Power Ratio (MCPR) safety limit is defined 
in the Bases to Technical Specification 2.1.1 as that limit which 
``ensures that during normal operation and during Anticipated 
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in 
the core do not experience transition boiling.'' The MCPR safety 
limit is re-evaluated for each reload and, for GGNS Cycle 11, the 
analyses have concluded that a two-loop MCPR safety limit of 1.09, 
based on the application of GE's [General Electric's] NRC-approved 
cycle-specific MCPR safety limit methodology demonstrates that this 
acceptance criterion is satisfied. For single-loop operation, a MCPR 
safety limit of 1.10, based on GE's [NRC-approved cycle-specific 
MCPR safety limit methodology, also demonstrates that this 
acceptance criterion is satisfied. Core MCPR operating limits are 
developed to support the Technical Specification 3.2 requirements 
and ensure these safety limits are maintained in the event of the 
worst-case transient. Since the MCPR safety limit will be maintained 
at all times, operation under the proposed changes will ensure at 
least 99.9% of the fuel rods in the core do not experience 
transition boiling. Therefore, these changes to the Minimum Critical 
Power Ratio (MCPR) safety limit do not affect the probability or 
consequences of an accident.
    GE's NRC-approved GESTAR-II cycle-specific MCPR safety limit 
methodology has been applied and has no effect on the probability or 
consequences of any accidents previously evaluated. As previously 
licensed, one exception to GESTAR is that the mis-oriented and mis-
located bundle events will continue to be analyzed as accidents 
subject to the acceptance criteria in the current licensing basis. 
The design of the GE11 fuel bundles is such that the bundles are not 
likely to be mis-oriented or mis-located and the normal 
administrative controls will be in effect for assuring proper 
orientation and location. Therefore, the probability of a fuel 
loading error is not increased. This analysis ensures that 
postulated dose releases will not exceed a small fraction (10 
percent) of 10CFR100 limits. Therefore, the probability or 
consequences of accidents previously evaluated are unchanged.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The GE11 fuel to be used in Cycle 11 is of a design compatible 
with fuel present in the core and used in the previous cycle. 
Therefore, the GE11 fuel will not create the possibility of a new or 
different kind of accident. The proposed changes do not involve any 
new modes of operation, any changes to setpoints, or any plant 
modifications. The proposed revised MCPR safety limits have been 
shown to be acceptable for Cycle 11 operation. Compliance with the 
applicable criterion for incipient boiling transition continues to 
be ensured. The proposed MCPR safety limits do not result in the 
creation of any new precursors to an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.

[[Page 46435]]

    The MCPR safety limits have been evaluated in accordance with 
GE's NRC-approved cycle-specific methodology to ensure that during 
normal operation and during AOOs, at least 99.9% of the fuel rods in 
the core are not expected to experience transition boiling. One 
exception to GESTAR is that the mis-oriented and mis-located bundle 
events will continue to be analyzed as accidents subject to the 
acceptance criteria in the current licensing basis. This analysis 
ensures that postulated dose releases for the worst case mis-
oriented and mis-located bundle will not exceed a small fraction (10 
percent) of 10CFR100 limits. On this basis, the implementation of 
this GE methodology 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.
    Local Public Document Room Location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: June 23, 1999.
    Description of amendment request: The requested Technical 
Specification changes would revise those specifications associated with 
various engineered safety feature systems, which need no longer be 
credited following a design-basis fuel handling accident. The proposed 
changes affect conditions where irradiated fuel is handled in the 
primary or secondary containment, and also affect certain 
specifications related to performing core alterations. These changes 
are based on the revised analysis of the design-basis fuel handling 
accident for the Grand Gulf Nuclear Station. This requested change is 
consistent with the changes approved for the Perry Nuclear Power Plant 
Operating License (Amendment 102), and the industry-proposed change to 
the Technical Specification NUREGs, TSTF-51.
    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 significantly increase the 
probability or consequences of an accident previously evaluated.
    A new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis. Because the equipment affected by the revised operational 
conditions is not considered an initiator to any previously analyzed 
accident, inoperability of the equipment cannot increase the 
probability of any previously evaluated accident. The proposed 
requirements bound the conditions of the current design basis fuel 
handling accident analysis which concludes that the radiological 
consequences are within the acceptance criteria of NUREG 0800, 
Section 15.7.4 and General Design Criteria 19. Therefore, the 
proposed changes do not significantly increase the probability or 
consequences of any previously evaluated accident.
    Removing a one time only allowance granted by Amendment 129 to 
the Operating License that is no longer in affect is an 
administrative change. Therefore, the proposed change does not 
significantly increase the probability or consequences of any 
previously evaluated accident.
    Based on the above, neither the proposed changes to the 
Technical Specifications nor that to the Operating License 
significantly increase the probability or consequences of any 
accident previously evaluated.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous analyzed.
    The new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis. The proposed changes do not introduce any new modes of 
plant operation and do not involve physical modifications to the 
plant. Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any previous analyzed.
    Removing a one time only allowance granted by Amendment 129 to 
the Operating License that is no longer in affect is an 
administrative change. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previous analyzed.
    Based on the above, neither the proposed changes to the 
Technical Specifications nor that to the Operating License create 
the possibility of a new or different kind of accident from any 
accident previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis and are established such that the radiological consequences 
are at or below the current GGNS [Grand Gulf Nuclear Station] 
licensing limit. Safety margins and analytical conservatisms have 
been evaluated and are well understood. Substantial margins are 
retained to ensure that the analysis adequately bounds all 
postulated event scenarios. The proposed change only eliminates the 
unnecessary margin from the analysis. The current margin of safety 
is retained.
    Specifically, the margin of safety for the fuel handling 
accident is the difference between the 10CFR100 limits and the 
licensing limit defined by NUREG 0800, Section 15.7.4. With respect 
to the control room personnel doses, the margin of safety is the 
difference between the 10CFR100 limits and the licensing limit 
defined by 10CFR50, Appendix A, Criterion 19 (GDC 19). The 
additional margin between the calculated doses for the postulated 
events and the corresponding licensing limit provides no useful 
purpose.
    The proposed applicability continues to ensure that the whole-
body and thyroid doses at both the control room and the exclusion 
area and low population zone boundaries are at or below the 
corresponding licensing limit. The margin of safety is unchanged; 
therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Removing a one time only allowance granted by Amendment 129 to 
the Operating License that is no longer in affect is an 
administrative change. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    Based on the above, neither the proposed changes to the 
Technical Specifications nor that to the Operating License result in 
a significant reduction in a margin of safety.
    Based on the above evaluation, operation in accordance with the 
proposed amendment involves no significant hazards considerations.

    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.
    Local Public Document Room Location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 6, 1998.

[[Page 46436]]

    Description of amendment request: The proposed change modifies the 
requirement to perform a Moderator Temperature Coefficient (MTC) test 
near the end of each cycle. This request constitutes a lead-plant 
submittal, submitted by Waterford 3 on behalf of the Combustion 
Engineering Owners Group (CEOG). CE NPSD-911, Amendment 1, ``Analysis 
of Moderator Temperature Coefficients in Support of a Change in the 
Technical Specifications End of Cycle Negative MTC Limit'' dated 
January, 1998 is provided as an Attachment to the application. 
Specifically, the proposed change modifies Technical Specification (TS) 
4.1.1.3.2c by adding a provision that eliminates the need to determine 
the MTC upon reaching two-thirds of core burnup if the results of the 
MTC tests required in TS 4.1.1.3.2a and 4.1.1.3.2b are within a 
specified tolerance. In addition, some editorial changes are proposed 
and the Bases change is included to support the changes in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    Under the proposed change, compliance with the TS Limiting 
Condition for Operation is achieved through a surveillance program 
consisting of beginning-of-cycle (BOC) measurements, plant parameter 
monitoring, and end-of-cycle (EOC) MTC predictions. This change 
eliminates the performance of the 2/3 Cycle MTC Surveillance when 
the BOC MTC Surveillances are within a required tolerance of the 
design value.
    The probability and consequences of an accident previously 
evaluated will not be increased because this change does not modify 
any assumptions used in the input to the safety analyses. The 
current safety calculations will remain valid because the allowed 
range of MTC values will not change.
    The Combustion Engineering analysis CE NPSD-911 and CE NPSD-911 
Amendment 1, demonstrate that if the startup test program has 
established that the core is operating as intended, and if the 
isothermal temperature coefficients measured at zero power during 
the cycle startup program, and at power prior to 40 EFPD [Effective 
Full Power Days], fall within the design value of plus or minus 
0.16 x 10-4 delta k/k/ deg.F, then the end-of-cycle best 
estimate prediction will also be within plus or minus 
0.16 x 10-4 delta k/k/ deg.F of true MTC.
    Removing the footnote that was applicable during Cycle 7 and 
providing a plus/minus for SR 4.1.1.3.2c is purely an administrative 
change.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    Plant operation and plant parameter TS limits will remain 
unchanged. There are no new changes in plant design nor are any new 
failure modes introduced. CE NPSD-911 analysis determined that if 
the MTC at the beginning-of-cycle is within plus or minus 
0.16 x 10-4 delta k/k/ deg.F of the design value then the 
MTC at the end-of-cycle will also be within plus or minus 
0.16 x 10-4 delta k/k/ deg.F of the design value.
    Removing the footnote that was applicable during Cycle 7 and 
providing a plus/minus for SR 4.1.1.3.2c is purely an administrative 
change.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The margin of safety will not be reduced because the range of 
allowed temperature coefficients will not be changed. The 
surveillance program consisting of beginning-of-cycle measurements, 
plant parameter monitoring, and end-of-cycle MTC predictions will 
ensure that the MTC remains within the range of acceptable values.
    Removing the footnote that was applicable during Cycle 7 and 
providing a plus/minus for SR 4.1.1.3.2c is purely an administrative 
change.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
    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, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: July 26, 1999.
    Description of amendment request: The proposed amendment would make 
the following line-item Technical Specification (TS) improvements:
    (1) Relocate TS Section 3/4.3.3.2, Instrumentation--Incore 
Detectors; TS 3/4.3.3.9, Instrumentation--Waste Gas System Oxygen 
Monitor; and TS 3/4.4.7, Reactor Coolant System `` Chemistry, to the 
Updated Safety Analysis Report (USAR) Technical Requirements Manual 
(TRM);
    (2) Change to TS 3/4.11.2, Radioactive Effluents--Explosive Gas 
Mixture, and TS Bases 3/4.11.2, Explosive Gas Mixture, to reflect the 
above proposed relocation of TS 3/4.3.3.9;
    (3) Revise the requirements of TS 3/4.4.6.1, Reactor Coolant System 
Leakage--Leakage Detection Systems, to require one monitor (gaseous or 
particulate) of the containment atmosphere radioactivity monitoring 
systems to be operable, rather than requiring both systems to be 
operable simultaneously; and
    (4) Revise the requirements of TS 3/4.3.3.1, Radiation Monitoring 
Instrumentation, to be consistent with the above proposed revision to 
TS 3/4.4.6.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit Number 1, in accordance with these 
changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiator, 
conditions or assumptions are affected by the proposed revisions to 
Technical Specification (TS) 3/4.3.3.1, Radiation Monitoring 
Instrumentation, TS 3/4.3.3.2, Instrumentation--Incore Detectors; TS 
3/4.3.3.9, Instrumentation--Waste Gas System Oxygen Monitor; TS 3/
4.4.7, Reactor Coolant System--Chemistry; TS 3/4.11.2, Radioactive 
Effluents--Explosive Gas Mixture; and TS 3/4.4.6.1, Reactor Coolant 
System Leakage--Leakage Detection Systems, and their associated TS 
Bases.
    The requirements of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7 
are proposed to be relocated from the TS to the DBNPS Updated Safety 
Analysis Report (USAR) Technical

[[Page 46437]]

Requirements Manual (TRM). These requirements would be relocated 
generally intact to the TRM whereby future changes would be subject 
to the regulatory controls of 10 CFR 50.59. These relocations are 
consistent with the NRC guidance provided in Generic Letter (GL) 95-
10, ``Relocation of Selected Technical Specifications Requirements 
Related to Instrumentation,'' or NUREG-1430, Revision 1, ``Standard 
Technical Specifications--Babcock and Wilcox Plants,'' dated April 
1995.
    The proposed revision to TS 3/4.11.2, Radioactive Effluents--
Explosive Gas Mixture, and its Bases is an administration change to 
a reference necessitated by the proposed relocation of TS 3/4.3.3.9 
to the USAR TRM.
    The proposed revision to TS 3/4.3.3.1 and TS 3/4.4.6.1 regarding 
the number of Reactor Coolant System (RCS) leakage detection 
monitors required and their allowed outage times is based upon the 
NRC's guidance of NUREG-1430, Revision 1. This proposed revision 
affects the TS only and does not reduce the number, diversity, or 
sensitivity of Reactor Coolant System leakage detection systems 
inside the containment building or as committed to in the DBNPS 
USAR.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident condition or 
assumption is affected by the proposed revisions. As described 
above, the revisions are consistent with the guidance of NRC GL 95-
10 or NUREG-1430, Revision 1. The proposed revisions, as described 
above, do not alter the source term, containment isolation, or 
allowable releases. The proposed changes, therefore, will not 
increase the radiological consequences of a previously evaluated 
accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed TS 
revisions. No new accident scenarios, transient precursors, failure 
mechanisms, or limiting failures are introduced as a result of the 
proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because the proposed revisions do not reduce or adversely affect the 
capabilities of any plant structures, systems or components. The 
proposed relocation of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7 to 
the USAR TRM is essentially an administrative change to the location 
and process by which these requirements are controlled and revised. 
Future revisions to these requirements relocated to the USAR TRM 
will be subject to the regulatory controls of 10 CFR 50.59. 
Therefore, these revisions will not result in a significant 
reduction in a margin of safety.
    The proposed revision to TS 3/4.11.2 and its Bases is 
administrative and reflects the relocation of TS 3/4.3.3.9 to the 
USAR TRM. Therefore, this revision will not result in a significant 
reduction in a margin of safety.
    The proposed revisions to TS 3/4.3.3.1 and TS 3/4.4.6.1 affect 
the number of containment atmosphere radioactivity monitors required 
by TS to be operable simultaneously. However, redundancy and 
diversity requirements are maintained in the TS for detecting 
Reactor Coolant System leakage. Although TS-allowed outage times are 
proposed to be increased consistent with NUREG-1430, Revision 1 
guidance, related compensatory action requirements are also being 
increased. Furthermore, the DBNPS commitments made for complying 
with Regulatory Guide 1.45, May, 1973, ``Reactor Coolant Pressure 
Boundary Leakage Detection Systems,'' are not changed by the 
proposed revisions. Along with the applicable revised TS 
requirements, 10 CFR 50, Appendix B, Criterion XVI will require 
prompt corrective action for inoperable leakage detection systems. 
Accordingly, these proposed revisions will not result in a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: July 26, 1999.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications to adopt the performance-based 10 
CFR Part 50, Appendix J, Option B approach for Type B and C containment 
leakage rate testing, and to relocate certain details of the tests into 
a Containment Leakage Testing Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit No. 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because accident initiators, 
conditions, or assumptions are not affected by the proposed changes.
    The proposed changes to the Technical Specifications and Bases 
implement 10 CFR [Part] 50 Appendix J Option B for Type B and C 
Local Leak Rate Testing, based on the guidance of Regulatory Guide 
1.163,
    ``Performance-Based Containment Leak-Test Program.'' Provided 
that components have performed satisfactorily on a historical basis, 
this guidance permits the use of extended testing frequencies. These 
proposed changes do not affect accident initiators, conditions, or 
assumptions.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
change the source term or total allowable releases. With the 
exception of the proposed increase in the containment air lock 
leakage limits, the proposed changes do not affect the total 
allowable containment leakage rates presently specified in the 
Technical Specifications. Although the air lock leakage limits are 
proposed to be increased, the accident analyses are based on the 
current TS allowable maximum bypass leakage, which is not proposed 
to be changed. Therefore, increases in leakage limits for individual 
components, such as the air locks and their door seals, which are 
constituents of bypass leakage, will have no effect on the 
radiological consequences described in the accident analyses.
    The proposed TS changes relating to implementation of 10 CFR 
[Part] 50 Appendix J Option B may result in a small, but acceptable 
increase in post-accident containment leakage, due to the increased 
probability that due to generally increased intervals between tests, 
an unacceptable leakage rate could go undetected for a longer length 
of time. NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' September, 1995, which provided the technical basis for 
the 10 CFR [Part] 50 Appendix J Option B rulemaking, provides a 
detailed evaluation of the expected leakage and its consequences and 
concludes that increased test frequencies are workable without 
significant risk impacts.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes. The proposed changes do not affect the methodology used in 
conducting containment leak rate testing. The proposed changes do 
not involve a change to the plant design or operation and, 
therefore, will not introduce any new or different failure modes or 
initiators.
    3. Not involve a significant reduction in a margin of safety.
    The proposed changes relating to implementation of 10 CFR [Part] 
50, Appendix J, Option B do not significantly affect the allowable 
containment leakage rates presently specified in the Technical 
Specifications. The Technical Specifications, under the proposed 
changes, will continue to ensure containment reliability by periodic 
testing performed in full compliance with 10 CFR [Part] 50, Appendix 
J.


[[Page 46438]]


    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.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: July 28, 1999.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.7.5.1, ``Ultimate Heat 
Sink,'' to allow operation on Modes 1 through 4 with an Ultimate Heat 
Sink water temperature of less than or equal to 90 deg.F, instead of 
the current limit of less than or equal to 85 deg.F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit No. 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions, or assumptions are significantly affected by the 
proposed change. The proposed change would increase the allowable 
Ultimate Heat Sink (UHS) water temperature, as specified in TS LCO 
3.7.5.1.b, from less than or equal to 85 deg.F to less than or equal 
to 90 deg.F. This water is used by the Service Water System to 
provide cooling to equipment that is used to mitigate accidents such 
as a Large Break Loss of Coolant Accident. This increase in Service 
Water temperature has been evaluated and the proposed change does 
not result in the operation of equipment important to safety outside 
their acceptable operating ranges.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed change does not 
change the source term, containment isolation, or allowable 
releases. The proposed increase in the Service Water System 
temperature has been evaluated with respect to the containment and 
equipment used to mitigate the consequences of accidents previously 
evaluated. These evaluations have determined that there are no 
significant increases in consequences.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
5 deg.F increase in UHS temperature. The proposed change does not 
result in installed equipment being operated outside their design 
operating ranges. No new or different equipment failure modes or 
mechanisms are introduced by the proposed change.
    3. Not involve a significant reduction in a margin of safety 
because the proposed 5 deg.F increase in UHS temperature does not 
result in significant changes to the initial conditions contributing 
to accident severity or consequences.

    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.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037 .
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 17, 1999.
    Description of amendment request: The proposed amendment modifies 
multiple surveillance requirements to support implementation of a 24-
month operating cycle.
    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 amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

A. Frequency Extensions

    The proposed Technical Specification (TS) changes involve a 
change in the surveillance testing intervals to facilitate a change 
in the Perry Nuclear Power Plant (PNPP) operating cycle from 18 
months to 24 months. The proposed TS changes do not physically 
impact the plant, nor do they impact any design or functional 
requirements of the associated systems. That is, the proposed TS 
changes do not degrade the performance of, or increase the 
challenges to, any safety systems assumed to function in the 
accident analysis. The proposed TS changes do not impact the TS 
surveillance requirements themselves, or the way in which the 
surveillances are performed. In addition, the proposed TS changes do 
not introduce any accident initiators, since no accidents previously 
evaluated have, as their initiators, anything related to the 
frequency of surveillance testing. Also, evaluation of the proposed 
TS changes demonstrated that the availability of equipment and 
systems required to prevent or mitigate the radiological 
consequences of an accident are not significantly affected because 
of other, more frequent testing that is performed, the availability 
of redundant systems and equipment, or the high reliability of the 
equipment. Since the impact on the systems is minimal, it is 
concluded that the overall impact on the plant accident analysis is 
negligible. Furthermore, a historical review of surveillance test 
results and associated maintenance records indicated that there was 
no evidence of any failures that would invalidate the above 
conclusions. Therefore, the proposed TS changes do not significantly 
increase the probability or consequences of an accident previously 
evaluated.

B. Allowable Value Changes

    The proposed changes in Allowable Values for the instrumentation 
include in Table 3.3.8.1-1 Items d and e of the Technical 
Specifications are the result of application of the Perry Instrument 
Setpoint Methodology (ISM) using plant specific drift values. 
Application of this methodology results in Allowable Values which 
more accurately reflect total instrumentation loop accuracy as well 
as that of test equipment and calculated drift between 
surveillances. The proposed changes will not result in any hardware 
changes. The instrumentation is not assumed to be an initiator of 
any analyzed event. Existing operating margin between plant 
conditions and actual plant setpoints is not significantly reduced 
due to these changes. The role of the instrumentation is in 
mitigating and thereby limiting the consequences of accidents. The 
Allowable Values have been developed to ensure that the design and 
safety analysis limits will be satisfied. The methodology used for 
the development of the Allowable Values ensures the affected 
instrumentation remains capable of mitigating design basis events as 
described in the safety analyses and that the results and 
radiological consequences described in the safety analyses remain 
bounding. Additionally, the proposed change does not alter the 
plant's ability to detect and mitigate events. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

[[Page 46439]]

C. Frequency Reductions to Semiannual

    The proposed Technical Specification (TS) changes involve a 
change in the surveillance testing intervals from 18 months to 
either 6 months or quarterly. The shorter frequencies are based on 
PNPP specific results of setpoint drift evaluations. The proposed 
more restrictive TS changes do not physically impact the plant, nor 
do they impact any design or functional requirements of the 
associated systems. That is, the proposed TS changes do not degrade 
the performance of, or increase the challenges to, any safety 
systems assumed to function in the accident analysis. The proposed 
TS changes do not impact the TS surveillance requirements 
themselves, or the way in which the surveillances are performed. In 
addition, the proposed TS changes do not introduce any accident 
initiators, since no accidents previously evaluated have, as their 
initiators, anything related to the frequency of surveillance 
testing. The proposed TS frequencies will demonstrate that the 
equipment and systems required to prevent or mitigate the 
radiological consequences of an accident are continuing to meet the 
assumptions of the setpoint evaluation, on a more frequent basis. 
Since the impact on the systems is minimal, and the assumptions of 
the safety analyses will be maintained, it is concluded that the 
overall impact on the plant accident analysis is negligible. 
Furthermore, a historical review of surveillance test results and 
associated maintenance records indicated that there was no evidence 
of any failures that would invalidate the proposed test frequencies. 
Therefore, the proposed TS changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed amendment would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

A. Frequency Extensions

    The proposed TS changes involve a change in the surveillance 
testing intervals to facilitate a change in the PNPP operating cycle 
length. The proposed TS changes do not introduce any failure 
mechanisms of a different type than those previously evaluated, 
since there are no physical changes being made to the facility. No 
new or different equipment is being installed. No installed 
equipment is being operated in a different manner. As a result, no 
new failure modes are being introduced. In addition, the 
surveillance test requirements themselves, and the way surveillance 
tests are performed, will remain unchanged. Furthermore, a 
historical review of surveillance test results and associated 
maintenance records indicated there was no evidence of any failures 
that would invalidate the above conclusions. Therefore, the proposed 
TS changes do not create the possibility of a new or different kind 
of accident from any previously evaluated.

B. Allowable Value Changes

    The proposed changes are the result of application of the ISM 
using plant specific drift values and do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. This is based on the fact that the method and manner of 
plant operation is unchanged. The use of the proposed Allowable 
Values does not impact safe operation of PNPP in that the safety 
analysis limits will be maintained. The propose Allowable Values 
involve no system additions or physical modifications to systems in 
the station. These Allowable Values were revised to ensure the 
affected instrumentation remains capable of mitigating accidents and 
transients. Plant equipment will not be operated in a manner 
different from previous operation, except that setpoints may be 
changed. Since operational methods remain unchanged and the 
operating parameters have been evaluated to maintain the station 
within existing design basis criteria, no different type of failure 
or accident is created.

C. Frequency Reductions to Semiannual or Quarterly

    The proposed TS changes involve a change in the surveillance 
testing interval due to the application of the ISM and plant 
specific drift analysis results. Also, the quarterly tests reflect 
current PNPP calibration practices, since the components are 
normally calibrated during the Channel Functional Test. The proposed 
TS changes do not introduce any failure mechanisms of a different 
type than those previously evaluated, since there are no physical 
changes being made to the facility. No new or different equipment is 
being installed. No installed equipment is being operated in a 
different manner. The proposed change does not impact core 
reactivity, or the manipulation of fuel bundles. As a result, no new 
failure modes are being introduced. In addition, the surveillance 
test requirements themselves, and the way surveillance tests are 
performed, will remain unchanged. Furthermore, a historical review 
of surveillance test results and associated maintenance records 
indicated there was no evidence of any failures that would 
invalidate the above conclusions. Therefore, the proposed TS changes 
do not create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed amendment will not involve a significant reduction 
in a margin of safety.

A. Frequency Extensions

    Although the proposed TS changes will result in changes in the 
interval between surveillance tests, the impact, if any, on system 
availability is small, based on other, more frequent testing that is 
performed, or the existence of redundant systems and equipment, or 
overall system reliability. Evaluations have shown there is no 
evidence of time dependent failures that would impact the 
availability of the systems. The proposed change does not 
significantly impact the condition or performance of structures, 
systems, and components relied upon for accident mitigation. The 
proposed change does not significantly impact any safety analysis 
assumptions or results. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

B. Allowable Value Changes

    The proposed change does not involve a reduction in a margin of 
safety. The proposed changes have been developed using a methodology 
to ensure safety analysis limits are not exceeded. As such, this 
proposed change does not involve a significant reduction in a margin 
of safety.

C. Frequency Reductions to Semiannual or Quarterly

    The proposed TS changes will result in a shorter interval 
between surveillance tests to ensure that the assumptions of the 
safety analysis are maintained. The impact, if any, on system 
availability is small, as a result of this more frequent testing 
that is performed. The proposed change does not significantly impact 
the condition or performance of structures, systems, and components 
relied upon for accident mitigation. The proposed change does not 
significantly impact any safety analysis assumptions or results. 
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.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: August 4, 1999.
    Description of amendment request: The amendment would incorporate 
an additional option into the Required Actions for Technical 
Specification 3.9.1, ``Refueling Equipment Interlocks.'' The change 
would provide additional Required Actions when the refueling interlocks 
are inoperable. The alternative would permit continued refueling 
activities once control rod withdrawal is blocked and operators verify 
that all appropriate controls 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:

    1. The proposed change does not involve a significant increase 
in the probability or

[[Page 46440]]

consequences of an accident previously evaluated.
    The refueling interlocks are explicitly assumed in the Perry 
Nuclear Power Plant Updated Safety Analysis Report (USAR) analyses 
of the control rod removal error and fuel loading error during 
refueling. This analysis evaluates the probability and consequences 
of control rod withdrawal during refueling. Criticality and, 
therefore, subsequent prompt reactivity excursions are prevented 
during the loading of fuel, provided all required control rods are 
fully inserted. The refueling interlocks accomplish this by 
preventing loading fuel into the core with any control rod 
withdrawn, or by preventing withdrawal of a rod from the core during 
fuel loading. When the refueling interlocks are inoperable, the 
current method of preventing fuel loading when a control rod is 
withdrawn, is to prevent fuel movement. This method is currently 
required by the Technical Specifications. An alternate method to 
ensure that fuel is not loaded into a cell with the control rod 
withdrawn is to prevent control rods from being withdrawn and verify 
that all control rods required to be inserted are fully inserted. 
The proposed Technical Specification Required Actions will require 
that a control rod block be placed in effect, thereby ensuring that 
control rods are not subsequently inappropriately withdrawn. 
Additionally, following placing the control rod withdrawal block in 
effect, the proposed actions will require that all required control 
rods be verified to be fully inserted. This verification is in 
addition to the requirements to periodically verify control rod 
position by other Technical Specification requirements. These 
proposed actions will ensure that control rods are not withdrawn and 
cannot be inappropriately withdrawn, because an electrical or 
hydraulic block to control rod withdrawal is in place. Like the 
current requirements, the proposed will ensure that unacceptable 
operations are blocked (e.g., loading fuel into a cell with a 
control rod withdrawn, except when following the requirements of LCO 
3.10.6, ``Multiple Control Rod Removal--Refueling,'' which is 
unaffected by this change). The proposed additional Required Actions 
provide an equivalent level of assurance that fuel will not be 
loaded into a core cell with a control rod withdrawn as do the 
current Required Action or the Surveillance Requirement. Therefore, 
the proposed change does not significantly increase the probability 
or consequences of an accident previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change in the Technical Specification requirements does not 
involve a change in the plant design, or to the status of the 
reactor core during refueling. The proposed actions will ensure that 
control rods are not withdrawn and cannot be inappropriately 
withdrawn, because an electrical or hydraulic block to control rod 
withdrawal is in place. Although the exact method by which the 
control rod withdrawal block is inserted is revised, the net effect 
is equivalent. The requirements will continue to ensure that fuel is 
not loaded into the core when a control rod is withdrawn, except 
when following the requirements of LCO 3.10.6, ``Multiple Control 
Rod Removal--Refueling,'' which is unaffected by this change. 
Therefore, no new failure modes are introduced, and the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    As discussed in the Bases for the affected Technical 
Specification requirements, inadvertent criticality is prevented 
during the loading of fuel provided all required control rods are 
fully inserted during the fuel insertion. The refueling interlocks 
function to support the refueling procedures by preventing control 
rod withdrawal during fuel movement and the inadvertent loading of 
fuel when a control rod is withdrawn. The proposed change will allow 
the refueling interlocks to be inoperable and fuel movement to 
continue only if a control rod withdrawal block is in effect and all 
required control rods are verified to be fully inserted. These 
proposed Required Actions provide an equivalent level of protection 
as the refueling interlocks by preventing a configuration which 
could lead to an inadvertent criticality event. The refueling 
procedures will continue to be supported by the proposed Required 
Actions because control rods cannot be withdrawn and as a result 
fuel cannot be inadvertently loaded when a control rod is withdrawn, 
except when following the requirements of LCO 3.10.6, ``Multiple 
Control Rod Removal--Refueling,'' which is unaffected by this 
change. Therefore, the proposed changes do not cause 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.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: February 23, 1999.
    Description of amendment request: The proposed license amendment 
would remove redundant boron concentration monitoring requirements 
specified for operating Modes 3 through 6 by deleting Technical 
Specification 3/4.1.2.9, ``Reactivity Control Systems--Boron 
Dilution.'' These requirements were interim measures intended to apply 
until a permanent boron dilution alarm system was installed and 
functional.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment does not involve changes to previously 
evaluated accident initiators. The proposed deletion of the 
redundant boron concentration verification requirements do not 
impact the results of existing accident analyses, and will have no 
adverse impact on any plant system performance. TS 3/4.1.2.9 
provides mode and charging pump dependent monitoring requirements 
for RCS boron concentration that are designed to detect an unplanned 
boron dilution event in MODES 3 through 6 in the absence of an 
automatic alarm system, and is based on the time requirements for 
operator action specified in Section 15.4.6 of the Standard Review 
Plan (SRP). This specification evolved from interim measures that 
were proposed by FPL until the boron dilution alarm system (BDAS) 
could be made completely functional following initial start-up of 
St. Lucie Unit 2. The BDAS is completely functional and provides 
redundant control room alarms to alert operators to the occurrence 
of an unplanned boron dilution event in Modes 3 through 6. The alarm 
setpoints are based on Chemical and Volume Control System (CVCS) 
malfunction analyses, and satisfy the same SRP acceptance criteria 
upon which the monitoring requirements of TS 3/4.1.2.9 were based. 
Therefore, operation of the facility in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The amendment 
will remove requirements from the facility technical specifications 
that were proposed by FPL as interim measures until the boron 
dilution alarm system became completely functional. The amendment 
will not alter the design of St. Lucie plant systems described in 
the Updated Final Safety Analysis Report (UFSAR), and the plant 
configuration will continue to remain consistent with assumptions 
used in the existing accident analyses. Therefore, operation of the 
facility in accordance with the proposed amendment would not create 
the possibility of a new or different kind of

[[Page 46441]]

accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment has been evaluated with respect to 
the applicable safety analyses. The BDAS provides a continuous, 
early warning capability to detect a boron dilution event in Modes 
3, 4, 5 and 6, and satisfies the same SRP time requirements for 
operator action as the interim TS that is proposed for deletion. 
BDAS setpoints are determined and/or validated for each fuel cycle 
to ensure they remain consistent with the CVCS malfunction analyses 
of record, and changes that may become necessary are controlled 
pursuant to 10 CFR 50.59. The minimum required Shutdown Margin is 
not changed by this proposal. Therefore, operation of the facility 
in accordance with the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Sheri R. Peterson.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Dade County, Florida

    Date of amendment request: July 27, 1999.
    Description of amendment request: The proposed amendments request 
that Turkey Point Unit 3 Technical Specification (TS) 3/4.8.1, A.C. 
SOURCES,TS 3/4.4.3, PRESSURIZER, and TS 3/4.5.2, ECCS SUBSYSTEMS--
Tavg GREATER THAN OR EQUAL TO 350 deg.F, be revised on a 
one-time basis to extend the Allowed Outage Time (AOT) for an 
inoperable Emergency Diesel Generator (EDG) from 72 hours to 7 days. 
The proposed one-time AOT extension will be used to replace the Unit 3 
EDG engine radiators prior to the Spring 2000 refueling outage. 
However, replacement of the radiator is a very labor-intensive 
evolution that cannot be performed within the existing 72 hour AOT. The 
proposed AOT extension will allow the radiator replacement activity to 
be completed successfully in a safe manner. The extended AOT will be 
applied to one EDG at a time in a sequential manner. When the radiator 
replacement activity is complete on one engine, it will be returned to 
service so that work can proceed on the redundant EDG. It should be 
noted that although the proposed changes apply only to Unit 3, the Unit 
4 TSs are administratively affected since the TSs are combined for both 
units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The Emergency Diesel Generators (EDG) are part of the on-site 
electrical power distribution system. They function as a standby 
power source in the event that the preferred A.C. power supply, 
i.e., offsite power, is interrupted. While certain failures in the 
electrical distribution system can lead to a loss of offsite power 
which is a design basis event for the plant, the EDGs are not 
assumed to be an initiating condition of any accident evaluated in 
the safety analysis report. Therefore, a one-time extension in the 
EDG Allowed Outage Time (AOT) does not involve a significant 
increase in the probability of an accident previously evaluated.
    The purpose of the proposed license amendment is to permit on-
line replacement of the Unit 3 EDG radiators. The radiators are part 
of the closed-loop diesel engine cooling water system and do not 
interface with any system or component that contains radioactivity. 
The EDGs do supply A.C. power to the emergency core cooling and 
containment heat removal systems during accidents that involve loss 
of offsite power. However, no changes are predicted for the 
postulated post-accident releases since adequate EDG capacity will 
be available under the conditions of the proposed license amendment 
to accommodate any design basis accident condition. Accordingly, the 
consequences of accidents previously evaluated in the safety 
analysis report are not changed by an extended EDG outage.
    Probabilistic Safety Assessment (PSA) techniques were used to 
evaluate the impact of a one-time extension of the EDG AOT from 72 
hours to 7 days. The results of these analyses indicate that 
extending the AOT for the purpose of replacing the engine radiator 
cores represents an acceptably small impact on Core Damage 
Probability.
    Based on the above, FPL concludes that the proposed amendment 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed change does not alter the design, physical 
configuration, or modes of operation of the plant. Plant 
configurations that are prohibited by Technical Specifications will 
not be created by the one-time EDG AOT extension. Therefore, the 
proposed activity does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed license amendment will extend by 96 hours the 
requirement to shutdown the plant when a Unit 3 EDG is removed from 
service for maintenance. The one-time AOT extension will not alter 
plant equipment, setpoints, or operating practices that provide the 
existing margins of safety. Therefore, the 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: August 27, 1998.
    Description of amendment request: The amendment would delete the 
requirements for an emergency plan from the 10 CFR Part 50 license and 
technical specifications after the spent nuclear fuel is transferred to 
a Part 72 licensed independent spent fuel storage installation (ISFSI).
    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.
    The proposed elimination of the emergency plan requirements from 
the 10 CFR 50 license is predicated on completion of transfer of the 
spent nuclear fuel to the proposed 10 CFR 72 ISFSI licensed area and 
removal of the reactor vessel and internals from the 10 CFR 50 
licensed area of the site.

[[Page 46442]]

Removal of the potential radiological source terms for accidents 
previously evaluated effectively eliminates the credibility of the 
accidents, therefore, elimination of the emergency plan requirements 
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.
    The proposed change is deletion of emergency plan requirements 
and, as such, has no direct impact on plant equipment or the 
procedures for operating plant equipment. Therefore, it does 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.
    Following the removal of the spent nuclear fuel and the reactor 
vessel and internals from the 10 CFR 50 licensed area, the remaining 
credible accidents are limited to decommissioning activities. The 
potential accidents associated with decommissioning activities are 
presented in the TNP [Trojan Nuclear Plant] Decommissioning Plan and 
have been shown to have consequences less than the EPA PAGs 
[Environmental Protection Agency Protective Action Guidelines]. 
Following the removal of the spent nuclear fuel and the reactor 
vessel (including the internals) from the 10 CFR 50 site, no 
credible accidents associated with the remaining decommissioning 
activities would require pre-planned emergency measures to avoid 
acute radiation doses. The deletion of the Trojan Nuclear Plant 
Permanently Defueled Emergency Plan will not result in a reduction 
in the margin of safety previously analyzed. Therefore, the proposed 
10 CFR 50 license amendment 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.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensee: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Section Chief: Michael T. Masnik.

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

    Date of amendment request: February 19, 1998, as supplemented July 
28, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 Technical Specifications (TSs) proposes to revise 
the Radioactive Effluents Technical Specifications (RETS) in accordance 
with Generic Letter 89-01 (GL-89-01), to make changes to implement 
revised 10 CFR Part 20 requirements, and to make administrative changes 
under 10 CFR 50.36a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of any accident 
previously evaluated?
    A. The proposed changes involve (1) combining related LCO and 
surveillance requirements from Sections 2.0 and 3.0, respectively, 
of the Indian Point 3 (IP3) RETS and relocating this text to the new 
Radiological Effluent Controls (REC) section of the ODCM, (2) 
relocating the bases contained in Section 4.0 of the RETS to the 
ODCM REC, (3) relocating the detailed reporting requirements 
contained in Section 5.0 of the RETS to the ODCM REC, and (4) 
updating references to 10 CFR Part 20. Additional changes include 
formatting both the remaining RETS and the new REG to more closely 
model Standard Technical Specifications (STS), revising the 
frequency of the Radioactive Effluent Release Report in accordance 
with 10 CFR 50.36a, relocating all definitions to Appendix A of the 
Technical Specifications and adding/deleting definitions as 
necessary, and adding a new Special Reports section to the ODCM. 
Most of the changes are (1) consistent with the guidance provided in 
the generic letter, NUREG-1301, or provisions of 10 CFR; or (2) 
editorial. Editorial changes include the relocation of text, 
correction of typographical and punctuation errors, renumbering, 
reformatting, immaterial wording revisions/deletions/clarifications 
which do not change intent, and updating references.
    B. The proposed revisions to the liquid and gaseous release rate 
limits, the relocation of the old 10 CFR 20.106 requirements to the 
new 10 CFR 20.1302, and the revision to the TS bases for the Liquid 
Holdup Tank activity will involve no change in the types or amounts 
of effluents that will be released, nor will there be an increase in 
individual or cumulative occupational radiation exposures.
    The changes of definitions, terminology, paragraph references, 
and report submittal frequency are necessary to keep IP3 TS 
consistent with revised federal regulations (i.e., 10 CFR 20 and 10 
CFR 50.36(a)). Record retention and reporting requirements will 
continue to meet NRC regulations. These changes are administrative 
in nature and do not affect plant hardware or operation.
    The changes do not impact the operation, design, configuration, 
or testing of plant structures, systems or components. As such, the 
proposed changes do not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any previously evaluated?
    A. The changes do not impact the operation, design, 
configuration, or testing of plant structures, systems or 
components. The changes do not result in a change in type or amount 
of radiological effluents released. As such, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    A. The changes are being made in accordance with NRC guidance 
and continue to assure compliance with the applicable regulatory 
requirements including 10 CFR 20. The changes do not result in a 
change in the types or amounts of effluents released. The current 
level of radiological effluent control will be maintained. As such, 
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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: March 18, 1996 (PA-192).
    Description of amendment request: The proposed amendment would 
update the Rancho Seco cask drop analysis and establish the cask drop 
event as the design-basis event for plant operation in the permanently 
defueled mode. The proposed amendment would also make editorial changes 
to the Permanently Defueled Technical Specifications and Bases by 
adding the word ``heavy'' to specification D3.3 and eliminating 
references to the MP-187 cask in specification D3.3 and D4.3.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 46443]]


    The District has reviewed the proposed changes against each of 
the criteria in 10 CFR 50.92, and, based on the above safety 
analysis, concludes:
    Using the Gantry Crane to handle a fully loaded transfer cask in 
the Fuel Storage Building will not create a significant increase in 
the probability or consequences of an accident previously evaluated 
in the SAR [safety analysis report], because the conservative dose 
consequence calculated for the updated, design basis cask drop event 
resulted in an exposure (224 mrem) that is:
    1. A very small percentage ([approximately] 0.9%) of the 10 CFR 
100 design basis accident dose limit of 25 rem total body;
    2. A small percentage ([approximately] 3.6%) of the NUREG-0612 
control of heavy loads accident dose limit of 6.25 rem total body;
    3. Well within ([approximately] 4.5%) of the old EPA 
[Environmental Protection Agency] NUREG-0654 plume exposure 
Protective Action Guidelines of 500 mrem total body dose;
    4. Well within the new EPA 1 to 5 rem Total Effective Dose 
Equivalent (TEDE) Protective Action Guidelines (PAGs) specified in 
document EPA-400-R-92-001, Table 2-1, May 1992;
    5. Less than the maximum hypothetical Rancho Seco Independent 
Spent Fuel Storage Installation design basis accident (375 mrem 
total body dose);
    6. Less than the original Rancho Seco operating design basis for 
the Fuel Storage Building FHA [fuel-handling accident] exposure (399 
mrem);
    7. Less than the original Rancho Seco operating design basis for 
the Reactor Building FHA exposure (477 mrem); and
    8. Much less than the original Rancho Seco operating design 
basis Maximum Hypothetical Accident exposure (3,600 mrem).
    Therefore, the conservatively calculated 224 mrem cask drop 
design basis accident exposure is (1) relatively small and (2) not 
considered a significant hazard.
    Also, the probability of occurrence of the FHA, which is the 
current design basis accident, is similar to the probability of 
occurrence of the updated cask drop event. The FHA is assumed to 
occur because the fuel handling bridge is not single failure proof. 
Likewise for the cask drop scenario, since the Gantry Crane is not 
single failure proof, this Safety Analysis Report evaluates the 
Gantry Crane dropping a loaded spent fuel cask.
    This Safety Analysis Report analyzes the dropped cask accident 
scenario even though the Gantry Crane and fuel handling bridge are:
    1. Designed to safely handle their respective loads (i.e., a 
loaded transfer cask and a spent fuel assembly, respectively; and
    2. In compliance with the design and administrative requirements 
addressed in NUREG-0612, ``Control of Heavy Loads at Nuclear Power 
Plants.''
    A loaded cask transfer drop is a very unlikely event because of 
the numerous Gantry Crane safety features described in the above 
safety Analysis Report. These features described above include:
    1. Gantry Crane Administrative Safety Features;
    2. Gantry Crane Design Safety Features;
    3. General Gantry Crane Control System Design Safety Features;
    4. Gantry Crane Radio Control System Design Safety Features;
    5. Hoist Design Safety Features; and
    6. Trolly and Bridge Design Safety Features.
    The updated cask drop accident scenario will not create the 
possibility of a new or different type of accident than previously 
evaluated in the SAR, because the DSAR [defueled SAR] currently 
evaluates a cask drop event. The cask drop scenario evaluated in the 
above Safety Analysis Report just updates the existing cask drop 
analysis. The updated cask drop analysis only:
    1. Identifies the type of spent fuel cask that Rancho Seco will 
use;
    2. Results in a change to the calculated dose consequence 
associated with the current, bounding, design basis accident (i.e., 
the FHA); and
    3. Results in a change to the existing Rancho Seco cask drop 
analysis.
    The updated, design basis, cask drop event will not involve a 
significant reduction in the margin of safety, because the 
conservatively calculated dose consequence associated with the 
postulated drop of a spent fuel transfer cask is:
    1. Relatively small (i.e., 224 mrem) compared to the eight 
accident limits and previously calculated accident doses summarized 
above;
    2. A very unlikely event;
    3. Not a significant hazard; and
    4. Not a public health and safety concern.
    This conclusion is the same for the FHA, which is the current, 
bounding, Rancho Seco design basis accident.
    Also, the Emergency Planning Zone remains unchanged for this 
updated, cask drop accident scenario. No significant changes to the 
Rancho Seco Emergency Plan result from this proposed change to the 
updated, design basis accident at Rancho Seco.

    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. The staff also reviewed the proposed editorial changes for 
no significant hazards consideration. The proposed editorial changes do 
not affect the design or operation of the facility and also satisfy the 
three standards of 10 CFR 50.92(c). Therefore, the NRC staff proposes 
to determine that the requested amendment involves no significant 
hazards consideration.
    Local Public Document Room location: Central Library, Government 
Documents, 828 I Street, Sacramento, California 95814
    Attorney for licensee: Dana Appling, Esq., Sacramento Municipal 
Utility District, P.O. Box 15830, Sacramento, California 95852-1830
    NRC Section Chief: Michael T. Masnik

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: March 12, 1998, as supplemented April 
24, August 20 and November 20, 1998, and February 3, 1999
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) of each unit to conform with 
NUREG-1431, Revision 1, ``Standard Technical Specifications--
Westinghouse Plants.'' The Commission had previously issued a Notice of 
Consideration of Issuance of Amendments in the Federal Register on May 
25, 1999, (64 FR 28218) covering all the proposed changes that were 
within the scope of NUREG-1431. The following descriptions and no 
significant hazard analyses cover only those items that are beyond the 
scope of NUREG-1431. Associated with each change are administrative/
editorial changes which would make the new or revised requirements fit 
into the format of NUREG-1431.
    1. The Standard Technical Specification (STS) terms FQW(Z) and 
FQC(Z) in Limiting Condition for Operation (LCO) 3.2.1 would be deleted 
and the terms FQ(Z), ``steady state'' limit and ``transient'' limit 
would be used. (Significant Hazards Evaluation A)
    2. The STS wording in Required Action 3.2.4.A to ``reduce'' thermal 
power would be revised to ``limit'' thermal power to allow entry into 
the LCO applicability during startup when QPTR may be in excess of 1.02 
due to transient core conditions which are usually self-correcting. (A)
    3. The Applicability of LCO 3.2.4 would be revised to be consistent 
with the Applicability for the AFD LCO to eliminate subtle differences 
between the two LCO Applications which were previously the same. (M)
    4. The Reactor Coolant System Loop Test specified in the TS LCO 3/
4.10.4 would not be included. (L-1)
    5. A new Action would be added to the Emergency Core Cooling System 
(ECCS)--Shutdown LCO 3.5.3. The new Action deals with the centrifugal 
charging subsystem. (L-2)
    6. The Reactor Coolant Pump (RCP) seal injection flow requirements 
of 3.5.5 would be revised. The requirement to verify a single operating 
point would be changed to require verification of a range of values on 
an operating curve. (M)
    7. The time allowed to reduce the power range neutron flux setpoint 
in 3.7.1 to within the required limit would be extended and made 
applicable in Mode 1 only. (L-3 and L-3a)

[[Page 46444]]

    8. The Actions in 3.7.2 for an inoperable Main Steam Isolation 
Valve (MSIV) would be revised to take credit for the redundant MSIVs in 
each steam line. (L-4)
    9. An Action would be added to the Service Water (SW) LCO 3.7.8 
that accounts for the redundant automatic turbine building isolation 
valves in each Farley SW train. (L-5)
    10. The diesel generator accelerated Test Table 3.8.1-1 would be 
deleted. (LA)
    11. The AC Sources--Shutdown surveillance 3.8.2.1 would be revised 
to more clearly state the required surveillances. (L-6 and L-6a)
    12. The Actions 3.8.4 and 3.8.9 for an inoperable SW intake 
structure Battery and Distribution System would be revised to more 
accurately reflect the Farley design. (L-7)
    13. The STS footnote to ESFAS Table 3.3.2-1 would be revised to be 
consistent with the design of the Farley main steam system. (L-8)
    14. A new Condition C would be added to LCO 3.3.4 to address 
actions associated with the source range neutron flux monitor. (M)
    15. LCO 3.3.5 would be revised to accommodate the addition of a 
degraded grid alarm function. (M)
    16. The specific title in 5.1.2 for the control room command 
function would be replaced with a more general description. (L-9)
    17. The specific title in 5.3.1 of Health Physics Supervisor would 
be replaced with a more general description. (A)
    18. The inspection frequency specified in 5.5.7 for the RCP 
flywheel would be revised to be consistent with the NRC-approved WCAP-
14535A, ``Topical Report on RCP Flywheel Inspection Elimination,'' 
November 1996. (L-10)
    19. The Health Physics Supervisor title in 5.7.1.c would be 
replaced with a more general description. (L-11)
    20. The Emergency Diesel General (DG) Failure Report in 5.6.7 would 
be revised to be consistent with the latest Farley commitments for DG 
failure tracking and reporting. (L-12)
    21. A note would be added to Surveillance Requirement (SR) 3.4.1.4 
that would not require this surveillance until 7 days after reaching 
greater than 90% power. (M)
    22. SR 3.4.5.2 would require verification that steam generator 
secondary side water levels are 74% (wide range). (M)
    23. LCO 3.4.15 would differ from the STS in several aspects. One 
aspect would extend the Allowable Outage Time from 7 days to 30 days 
for an inoperable leakage detection system. (L-13)
    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. Each proposed out-of-scope item described above is 
followed in parenthesis by either an A (for administrative changes), an 
M (for changes which would be more restrictive), an LA (for 
requirements that would be removed from the TS), or an L and a number 
(for changes that would be less restrictive). Following are the no 
significant hazards analyses corresponding to each of these 
designations.

[A--Administrative Changes]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve reformatting, renumbering, and 
rewording of the CTS. These changes involve no technical revisions 
to the CTS and were made to conform with the format and style of the 
STS. As such, these changes are administrative in nature and do not 
impact initiators of analyzed events or safety analyses assumptions 
relative to the mitigation of accidents or transient events. 
Therefore, these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in the methods governing normal plant operation. The 
proposed changes will not impose any new or different requirements 
or eliminate any existing requirements. In addition, the change does 
not alter assumptions made in the safety analyses and licensing 
basis. Therefore, the changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes are administrative in nature and do not 
involve any technical changes. As such, these changes do not impact 
any safety analysis assumptions and no question of safety is 
involved. Therefore, the changes do not involve a significant 
reduction in a margin of safety.

[M--More Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes provide more stringent requirements than 
previously existed in the CTS. These more stringent requirements are 
not assumed to be initiators of analyzed events and will not alter 
assumptions relative to mitigation of accident or transient events. 
The changes are evaluated to ensure no previously analyzed accident 
has been adversely affected. The more stringent requirements are 
imposed to ensure process variables, structures, systems and 
components are maintained consistent with the safety analyses and 
licensing basis. These changes will not alter assumptions relative 
to mitigation of an accident or transient event nor will they alter 
the operation of process variables, structures, systems, or 
components described in the safety analyses. Therefore, these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes add more restrictive requirements to the TS 
or make existing requirements more restrictive. The proposed changes 
do not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. The proposed changes do 
impose new or different requirements. However, these changes are 
consistent with assumptions made in the safety analysis and 
licensing basis. Thus, these changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes add more restrictive requirements to the TS 
or make existing requirements more restrictive and have been 
evaluated to ensure consistency with the safety analysis and 
licensing basis. As such, these changes do not impact any safety 
analyses assumptions and no question of safety is involved. 
Therefore, these changes do not involve a reduction in a margin of 
safety.

[LA--Removal of Requirements]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes relocate requirements from the CTS to a 
licensee controlled document. The document containing the relocated 
requirements will be maintained using the provisions of 10 CFR 
50.59. Therefore, the proposed changes will only reduce the level of 
regulatory control on these requirements. The level of regulatory 
control has no impact on the probability or the consequences of an 
accident previously evaluated. Thus, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes relocate requirements from the CTS to a 
licensee controlled document. The changes do not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any

[[Page 46445]]

existing requirements. The changes do not alter assumptions made in 
the safety analyses and licensing basis. Thus, the changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes relocate requirements from the CTS to a 
licensee controlled document for which future changes will be 
evaluated pursuant to the requirements of 10 CFR 50.59. The proposed 
changes do not reduce a margin of safety because they have no impact 
on any safety analysis assumptions. Therefore, these changes do not 
involve a significant reduction in a margin of safety.

[L-1--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves deleting the CTS 3/4.10.4, Reactor 
Coolant Loops Test Exception, requirements and does not result in 
any hardware changes. The proposed change deletes a test exception 
LCO that is no longer used or required at FNP. The natural 
circulation test, for which this exception is designed, was only 
required to be performed at FNP during the initial plant startup 
test program. The proposed changes do not impact the capability of 
the plant or any equipment to provide the required safety function 
as described in the FSAR. In addition, the results of the analyses 
described in the FSAR remain bounding. Also, the proposed changes do 
not impose any new safety analyses limits or alter the plants 
ability to detect and mitigate events. Therefore, this 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 involves changing the CTS requirements to 
delete a test exception that is no longer used and does not 
necessitate a physical alteration of the plant or changes in 
parameters governing normal plant operation. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change, which deletes CTS 3/4.10.4 does not involve 
a significant reduction in a margin of safety. The proposed change 
does not impact any safety analysis assumptions and does not impose 
any new safety analyses limits or alter the plants ability to detect 
and mitigate events. Therefore, the proposed change does not impact 
any margin of safety.

[L-2--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change does not result in any hardware changes. The ECCS 
components covered by this TS are not assumed to be initiators of 
any analyzed event. Therefore, this change does not involve a 
significant increase in the probability of an accident previously 
evaluated. The change would allow the required ECCS centrifugal 
charging subsystem to be inoperable for up to 72 hours providing the 
remaining operable ECCS components can provide the flow equivalent 
to a single operable train which will ensure 100% of the flow 
assumed in the safety analyses. Since the ability of the ECCS to 
perform its safety function is not lost, this change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not 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 will only more accurately define the 
minimum equipment required to be operable to perform the ECCS 
function while in this Condition. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change, which allows operation to continue for up 
to 72 hours with components inoperable in the required ECCS 
centrifugal charging subsystem, is acceptable based on the remaining 
ECCS components providing 100% of the required ECCS flow, the small 
probability of an event occurring in 72 hours that would require the 
ECCS, and the reduced potential for a unit transient resulting from 
the shutdown required by current TS for an inoperable required ECCS 
centrifugal charging subsystem. The proposed allowed outage time of 
72 hours for this condition is consistent with the time currently 
allowed for one train of ECCS to be inoperable in Modes 1-3. The 
exposure of the unit to the small probability of an event requiring 
ECCS during this time is insignificant and offset by the benefit 
gained through avoiding unnecessary plant transients. Therefore, 
this change does not involve a significant reduction in margin of 
safety.

[L-3--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes extend the time allowed to adjust the Power 
Range Neutron Flux-High trip setpoints for the case of two or more 
inoperable MSSVs per SG and/or positive Moderator Temperature 
Coefficient (MTC) and removes the requirement to adjust the Power 
Range Neutron Flux-High trip setpoints only one MSSV is inoperable 
and the MTC is zero or negative and do not result in any hardware or 
operating procedure changes. The affected trip setpoints, the 
requirement to reduce them or the time allowed to adjust them are 
not assumed to be an initiator of any analyzed event. In addition, 
the affected trip setpoints, the requirement to reduce them and the 
time allowed to adjust them are not a precursor to any accident 
analyses. Therefore, the proposed changes do not increase the 
probability of an accident previously evaluated. The Power Range 
Neutron Flux-High trip functions to mitigate the consequences of an 
analyzed event by shutting down the reactor. The proposed changes 
continue to provide assurance that the setpoints will be properly 
adjusted to ensure the system functions as assumed in the applicable 
safety analyses. Therefore, the consequences of an accident are not 
significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do 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 changes still ensure the operability of the trip function 
at the correct setpoint and will facilitate the adjustment of the 
setpoints such that the probability of error is minimized. Thus, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The time allowed to adjust the setpoints of the affected 
instrumentation is not a specific assumption of any safety analysis. 
For the case of a single inoperable MSSV with a zero or negative 
MTC, a reactor power reduction alone is sufficient to limit primary 
side heat generation such that overpressurization of the secondary 
side is precluded for any RCS heatup event. Furthermore, for this 
case there is sufficient total steam flow capacity provided by the 
turbine and the remaining OPERABLE MSSVs to preclude 
overpressurization in the event of an increased reactor power due to 
reactivity insertion, such as in the event of an uncontrolled RCCA 
bank withdrawal at power. The proposed changes still ensure the 
setpoints are reduced consistent with the assumptions of the safety 
analysis for the case of two or more inoperable MSSVs or a positive 
MTC. The proposed changes also reduce the potential for an 
inadvertent reactor trip that could result from adjusting the trip 
setpoints too quickly. As such, any reduction in a margin of safety 
will be insignificant and will likely be offset by the benefit 
gained from the reduced potential for an inadvertent plant trip.

[L3a--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change clarifies the Action requirement to reduce 
the power range neutron flux-high trip setpoint in Modes 2 and 3 and 
does not result in any hardware or operating procedure changes. The 
proposed change adds a note to the Action which specifies that the 
Action is only required in Mode 1. In Modes 2 and 3, other reactor 
trips (power range low and source range high) provide the required 
protection consistent with the acceptance criteria of the safety 
analysis. Therefore, the Action is not required in these Modes. The 
affected trip

[[Page 46446]]

setpoints are not assumed to be an initiator of any analyzed event. 
In addition, the affected trip setpoints are not a precursor to any 
accident analyses. Therefore, the proposed change does not increase 
the probability of an accident previously evaluated. The affected 
reactor trip functions mitigate the consequences of an analyzed 
event by shutting down the reactor. The proposed change continues to 
provide assurance that the required reactor trip functions operate 
as assumed in the applicable safety analyses. Therefore, the 
consequences of an accident are not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed 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 still ensures the operability of the 
reactor trip function at the correct setpoint for the correct Mode 
of operation. Thus, this change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change does not affect the ability of the MSSVs and 
reactor trip system to mitigate the applicable transients consistent 
with the assumptions of the safety analysis. The proposed change 
continues to ensure the acceptance criteria of the applicable safety 
analyses are met (primary and secondary system pressures are limited 
to within the required values). As such, any reduction in a margin 
of safety will be insignificant and will likely be offset by the 
benefit gained from the reduced potential for an inadvertent plant 
trip that could result from an error in adjusting the power range 
neutron flux-high trip setpoint (unnecessary in Mode 2).

[L-4--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change revises the Actions of the MSIV LCO in order 
to take credit for the redundant MSIV valves in each steam line. 
This change does not result in any hardware or operating procedure 
changes. The MSIVs are not assumed to be an initiator of any 
analyzed event and function to isolate the steam lines to mitigate 
analyzed events. As a result, the revision of this TS requirement 
does not affect the probability of an accident previously evaluated. 
The proposed change continues to provide adequate assurance that the 
MSIVs are either capable of performing their intended safety 
function or that the safety function has been performed (steam line 
isolated) or that power is reduced. The proposed change continues to 
limit plant operation when a single failure could prevent the 
isolation function from being accomplished. Therefore, the proposed 
change does not involve a significant increase in the consequences 
of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not 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 only affects the Actions of the MSIV 
LCO. The proposed change continues to ensure the MSIVs are either 
capable of isolating the steam lines or that the steam lines are 
isolated or power reduced. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change continues to ensure the MSIVs are either 
capable of isolating the steam lines or that the steam lines are 
isolated or power reduced. The proposed change continues to limit 
plant operation when a single failure could prevent the isolation 
function from being accomplished. Therefore, the proposed change 
also continues to preserve the assumptions of the applicable safety 
analyses. As such, the proposed change does not impact the 
assumptions of the applicable safety analyses. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

[L-5--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change revises the Actions of the SWS LCO in order 
to take credit for the redundant automatic turbine building 
isolation valves in each train of SWS. This change does not result 
in any hardware or operating procedure changes. The turbine building 
isolation valves are not assumed to be an initiator of any analyzed 
event and function to isolate the SWS flow to non-essential 
components. As a result, the revision of this TS requirement does 
not affect the probability of an accident previously evaluated. The 
proposed change continues to provide adequate assurance that the 
turbine building isolation valves are either capable of performing 
their intended safety function and accommodate a single failure or 
that the unit is placed in a condition where the function performed 
by these valves is no longer required. The proposed change continues 
to limit plant operation when a single failure could prevent the 
isolation function of these valves from being accomplished. 
Therefore, the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not 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 only affects the Actions of the SWS 
LCO. The proposed change continues to ensure the turbine building 
isolation valves are either capable of isolating the SWS system and 
accommodating a single failure or that the unit is placed in a 
condition where this isolation function is no longer required. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change continues to ensure the turbine building 
isolation valves are either capable of isolating the non-essential 
SWS loads and accommodating a single failure or that the unit is 
placed in a condition where the isolation function is no longer 
required. The proposed change continues to limit plant operation 
when a single failure could prevent the isolation function from 
being accomplished. Therefore, the proposed change also continues to 
preserve the assumptions of the applicable safety analyses. As such, 
the proposed change does not impact the assumptions of the 
applicable safety analyses. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

[L-6--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The elimination of the requirement to meet surveillance tests 
that verify functions which are not required in the Mode of 
applicability of this TS will not increase the probability of any 
accident previously evaluated. The proposed surveillance testing 
continues to provide adequate assurance of the operability of the 
required AC Source functions and therefore, does not involve an 
increase in the consequences of any 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 introduce a new mode of plant 
operation and does not involve a physical modification to the plant. 
Therefore, it does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    This change does not involve a significant reduction in a margin 
of safety since the operability of the required AC Source functions, 
continues to be determined in the same manner. Elimination of the 
surveillance test requirements for AC Source functions not required 
in these Modes does not impact the capability of the AC Sources to 
perform their safety function in these Modes.

[L6a--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The inclusion of a note consistent with the STS to provide an 
allowance not to perform certain surveillance tests on the AC Source 
required operable by the TS will not increase the probability of any 
accident previously evaluated. The required surveillance testing 
must still be performed (but not on the AC

[[Page 46447]]

Source while it is required operable by the TS) and will continue to 
provide adequate assurance of the operability of the required AC 
Source functions. Therefore, this change does not involve an 
increase in the consequences of any 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 introduce a new mode of plant 
operation and does not involve a physical modification to the plant. 
Therefore, it does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    This change does not involve a significant reduction in a margin 
of safety since the operability of the required AC Source functions, 
continues to be determined in the same manner. The allowance not to 
perform certain surveillance tests on the AC Source equipment when 
that equipment serves to meet the TS minimum required power source 
ensures a stable shutdown power supply to the unit and does not 
impact the capability of the AC Sources to perform their safety 
function in these Modes.

[L-7--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change effectively provides a longer allowed outage 
time for the Service Water Intake Structure (SWIS) DC distribution 
and battery systems. The proposed allowed outage time is consistent 
with the time allowed for a Service Water train to be inoperable. 
The DC power sources or their associated allowed outage times are 
not assumed to be initiators of any analyzed event. As such, the 
proposed change will not increase the probability of any accident 
previously evaluated. The appropriate required actions consistent 
with that for the equipment rendered inoperable must still be 
performed. The proposed actions will continue to provide adequate 
assurance of plant safety in the same manner as if the affected 
equipment were inoperable for reasons other than power availability. 
Therefore, this change does not involve an increase in the 
consequences of any 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 introduce a new mode of plant 
operation and does not involve a physical modification to the plant. 
Therefore, it does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    This change does not involve a significant reduction in a margin 
of safety since the inoperability of the SWIS distribution and 
battery systems affect only the Service Water system and the time 
allowed for restoration of an inoperable Service Water train remains 
unchanged. The allowance to declare the affected equipment 
inoperable and take the associated equipment TS actions continues to 
ensure plant safety by providing the same appropriate remedial 
measures for the affected equipment as would be applicable if that 
equipment were inoperable for reasons other than power availability. 
Therefore, the proposed change does not significantly impact any 
margin of safety.

[L-8--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves upgrading the ESFAS TS to more 
closely agree with the FNP design and safety analysis and does not 
result in any hardware changes. The proposed change revises the 
applicability for the initiating functions of the main steam line 
isolation function such that when a main steam line isolation valve 
is closed and the isolation function is accomplished, the automatic 
initiation of this function is no longer required operable. The 
ESFAS is not assumed to be an initiator of any analyzed event. The 
role of the ESFAS is in mitigating and thereby limiting the 
consequences of accidents. The proposed change continues to 
adequately ensure the operability of the ESFAS main steam line 
isolation function when the lines are unisolated and thereby ensures 
the protection provided by the function remains operable when 
required. Therefore, the results of the analyses described in the 
FSAR remain bounding. Additionally, the proposed changes do not 
impose any new safety analyses limits or alter the plants ability to 
detect and mitigate events. Therefore, this 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 involves upgrading the ESFAS TS to more 
closely agree with the FNP design and safety analysis and 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. Thus, this change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety? The proposed change, which upgrades the ESFAS TS to be 
more consistent with the FNP design and safety analysis does not 
involve a significant reduction in a margin of safety. The proposed 
change revises the Mode of applicability for the main steam line 
isolation ESFAS function. The proposed change continues to 
adequately ensure the operability of the isolation function when it 
is required and thereby ensures the protection provided by the 
function also remains available when required. As such, the results 
of the analyses described in the FSAR remain bounding and this 
change does not have a significant impact on any design basis safety 
analysis.

[L-9--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves changing the CTS administrative 
controls requirements regarding the Shift Supervisor (SS) 
responsibility to more closely agree with the STS requirements and 
does not result in any hardware changes. The requirement to issue 
annual directives regarding the SS responsibilities is deleted. The 
title Shift Supervisor is replaced with responsible SRO. In 
addition, an allowance for an RO (in Modes 5 and 6) to temporarily 
replace the SS is added. The proposed change also eliminates the 
specific restriction against the STA temporarily replacing the SS. 
The proposed changes do not impact the capability of the plant or 
any equipment to provide the required safety function as described 
in the FSAR. In addition, the results of the analyses described in 
the FSAR remain bounding. Additionally, the proposed changes do not 
impose any new safety analyses limits or alter the plants ability to 
detect and mitigate events. Therefore, this 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 involves changing the TS administrative 
controls regarding the responsibilities of the SS to more closely 
agree with the STS requirements and eliminates the title Shift 
Supervisor and does not necessitate a physical alteration of the 
plant or changes in parameters governing normal plant operation. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes, which revise the TS administrative 
controls requirements for SS responsibilities to be consistent with 
the STS requirements and eliminate the title Shift Supervisor do not 
involve a significant reduction in a margin of safety. The proposed 
changes do not impact any safety analysis assumptions and do not 
impose any new safety analyses limits or alter the plants ability to 
detect and mitigate events. Therefore, the proposed changes do not 
impact any margin of safety.

[L-10--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change affects only the interval allowed by the TS 
surveillance to perform RCP flywheel inspections. The time allowed 
between flywheel inspections is not specifically assumed to be a 
precursor or initiator of any analyzed event. The studies performed 
to justify the proposed time interval have shown it to be adequate 
to detect any flaws or degradation in the RCP flywheel. As such, the 
proposed change does not affect the probability of any initiating 
events assumed in the accident analyses. The proposed change will 
maintain an acceptable

[[Page 46448]]

level of safety by continuing to require RCP flywheel inspections at 
an interval shown to be adequate. Consequently, the proposed change 
will not have any affect on the consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different types of equipment will be installed) 
or changes in parameters governing normal plant operation. The 
proposed change only affects the interval allowed by the TS to 
inspect each RCP flywheel. The interval remains adequate to detect 
any degradation. Therefore, the possibility of a new or different 
kind of accident is not created by the proposed change.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change affects the interval allowed by the TS to 
inspect RCP flywheels. The proposed interval is based on the 
findings of WCAP-14535A and the associated NRC SER. The WCAP 
concludes that continued inspections of RCP flywheels are not 
necessary and overall plant safety could be increased by eliminating 
the inspections and reducing man rem dose as well as the potential 
for flywheel damage during disassembly and reassembly for 
inspection. The NRC SER requires the inspection of RCP flywheels be 
retained but the interval increased to once every 10 years. As such, 
the proposed change continues to conservatively assure the 
operability of the RCP flywheel while reducing man rem exposure and 
the potential for damage from disassembly and reassembly. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

[L-11--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves the revision of the term health 
physics supervisor to health physics supervision for the purpose of 
specifying the frequency of radiation surveillances in RWPs. The 
proposed change continues to provide adequate assurance that the 
radiation surveillances are performed within acceptable frequencies. 
The proposed change does not impact the capability of the plant or 
any equipment to provide the required safety function as described 
in the FSAR, or increase the potential radiation exposure of plant 
personnel. In addition, the results of the analyses described in the 
FSAR remain bounding. Additionally, the proposed change does not 
impose any new safety analyses limits or alter the plants ability to 
detect and mitigate events. Therefore, this 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 involves the supervisors who specify the 
radiation surveillance frequencies in high radiation areas and does 
not necessitate a physical alteration of the plant or changes in 
parameters governing normal plant operation. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change, which revises the TS requirements for the 
personnel who specify the frequencies of radiation surveillances in 
high radiation areas. The proposed change allows additional 
supervisory personnel to specify the required frequencies. The 
proposed change does not impact any safety analysis assumptions and 
does not impose any new safety analyses limits or alter the plants 
ability to detect and mitigate events. In addition, the proposed 
change continues to ensure adequate surveillances are performed in 
high radiation areas. Therefore, the proposed change does not impact 
any margin of safety.

[L-12--Less Restrictive]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves changing the CTS administrative 
controls requirements regarding the Emergency Diesel Generator (EDG) 
failure reporting requirement and does not result in any hardware 
changes. The proposed change potentially reduces the number of 
reports received by the NRC and revises the content to include valid 
failures and demands. The proposed change continues to provide 
adequate information to assess the EDG reliability at FNP. The 
proposed change does not impact the capability of the plant or any 
equipment to provide the required safety function as described in 
the FSAR. In addition, the results of the analyses described in the 
FSAR remain bounding. Additionally, the proposed change does not 
impose any new safety analyses limits or alter the plants ability to 
detect and mitigate events. Therefore, this 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 involves changing the TS administrative 
controls regarding the required EDG report to more closely agree 
with the STS requirements and does not necessitate a physical 
alteration of the plant or changes in parameters governing normal 
plant operation. Thus, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change, which revises the TS administrative 
controls requirement for an annual EDG report to be consistent with 
the STS requirement does not involve a significant reduction in a 
margin of safety. The proposed change does not impact any safety 
analysis assumptions and does not impose any new safety analyses 
limits or alter the plants ability to detect and mitigate events. In 
addition the proposed change continues to provide sufficient 
information to assess the reliability of the EDG at FNP. Therefore, 
the proposed change does not impact any margin of safety.

[L-13--Less Restrictive]

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    The proposed change extends the time allowed to restore an 
inoperable RCS leakage detection instrument to operable status. The 
CTS allow 7 days for restoration of the automatic RCS leak detection 
instrument and the proposed change would allow 30 days for 
restoration. However, adequate information continues to be furnished 
to the plant staff to assure that RCS leakage does not go 
undetected. In addition to the remaining operable automatic RCS leak 
detection instrument, the TS required actions provide remedial 
measures that ensure RCS leakage continues to be monitored by 
diverse means. As such, potential RCS leakage will not go undetected 
and operation with one required leak detection instrument inoperable 
continues to be limited by the TS. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not introduce any new equipment into 
the plant or alter the manner in which existing equipment will be 
operated. Therefore the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The applicable required actions and remaining operable leakage 
detection monitor provide adequate information to the plant staff to 
ensure that RCS leakage does not go undetected. In addition, 
operation with one required leak detection instrument inoperable 
continues to be limited by the TS (30 days). As such, potential RCS 
leakage will not go undetected and operation in the condition where 
a single failure could cause a loss of automatic leakage detection 
continues to be limited and therefore, this 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.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post

[[Page 46449]]

Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: July 29, 1999.
    Description of amendment request: The proposed amendments would 
change the Limiting Condition for Operation 3.1.7, ``Standby Liquid 
Control (SLC) System.'' The proposed amendments would change ``greater 
than the Region B limits,'' which could be misleading, to ``within the 
Region B limits.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to the Unit 1 and Unit 2 Technical 
Specifications do not increase the probability or consequences of 
any previously evaluated accident or transient. These changes are 
administrative in nature only and are intended to revise a 
misleading statement in Condition A of Limiting Condition for 
Operation (LCO) 3.1.7, ``Standby Liquid Control (SLC) System.'' The 
change ensures the proper condition is entered when expected and the 
sodium pentaborate solution temperature, concentration, and volume 
limits are not exceeded without appropriate actions being taken. As 
currently written, Condition A of LCO 3.1.7 could be entered 
whenever the sodium pentaborate solution is not within Region A 
limits, but is greater than Region B limits as depicted in Unit 1 
and Unit 2 Technical Specifications Figures 3.1.7-1 and 3.1.7-2. 
This is incorrect; Condition A should be entered whenever the 
solution is not within Region A limits, but is within Region B 
limits. If the solution is not within Region A limits and is greater 
than Region B limits, both Standby Liquid Control subsystems are 
inoperable and Condition C should be entered.
    Technical Specifications Figure 3.1.7-1 displays the sodium 
pentaborate solution volume versus concentration requirements; 
Figure 3.1.7-2 displays the solution concentration versus 
temperature requirements. Each figure contains three areas: Region 
A, Region B, and the area not in either Region A or Region B. Region 
A is the permissible region of continuous operation and is 
represented by a four- or five-sided area. Region B is the original 
licensing basis region and is represented by a four-sided area. If 
the sodium pentaborate solution temperature, concentration, and 
volume combinations are within Region A, the requirements of 10 CFR 
50.62, ``Requirements for reduction of risk from anticipated 
transients without scram (ATWS) events for light-water-cooled 
nuclear power plants,'' are met, no condition applies, and no 
actions need be taken. If solution temperature, concentration, and 
volume combinations are not within Region A, but within Region B, 
then the original licensing basis is met and operation within this 
region is acceptable for up to 72 hours (Unit 1 FSAR, section 3.8.4, 
Revision 6, page 3.8-6; Unit 2 FSAR, section 4.2.3.4.3, Revision 7, 
page 4.2-98). If solution temperature, concentration, and volume 
combinations are not within either region, then the ability of the 
Standby Liquid Control system to shut down the reactor is not 
assured and only eight hours is acceptable to restore the solution 
to at least within Region B before the plant must be shut down.
    Condition A contains misleading wording which could allow 
operation outside both Region A and Region B for more than eight 
hours. Specifically, it could be interpreted that Condition A allows 
the sodium pentaborate solution temperature, concentration, and 
volume to be greater than Region B limits for up to 72 hours. 
Because Region B is demarcated by a four-side area, the terms 
``within Region B'' and ``greater than Region B limits'' could be 
interpreted to indicate different, and mutually exclusive, areas of 
Figures 3.1.7-1 and 3.1.7-2. Indeed, ``greater than Region B 
limits'' could be interpreted to refer to most or all of the area 
neither in Region A nor Region B. For example, 20 weight percent 
sodium pentaborate solution at 50 deg.F is a point on Figure 3.1.7-2 
which is ``greater than the Region B limits,'' yet it is a point at 
which the solution will precipitate in the storage tank rendering 
the system incapable of injecting the proper amount of sodium 
pentaborate into the reactor pressure vessel. Obviously, both 
Standby Liquid Control subsystems would be inoperable if the 
solution were at this point and Condition C should be entered to 
limit severely the time the unit may continue to operate with the 
solution in this state. However, the wording of Condition A could 
cause an erroneous interpretation which would inappropriately extend 
this time from eight to 72 hours.
    The proposed changes correct the wording of Condition A to 
ensure this condition is not entered inappropriately and to ensure 
the proper condition is entered for those combinations of solution 
temperature, concentration, and volume not within Region A or Region 
B. These changes do not increase the probability of any previously 
evaluated accident or transient because they are administrative in 
nature and do not alter any plant operation or design features or 
requirements which could result in systems or components performing 
closer to their operational or design limits and thereby increasing 
the possibility of a failure. These changes do not increase the 
consequences of any previously evaluated accident or transient 
because they ensure the sodium pentaborate solution limits are not 
exceeded without appropriate actions being taken thereby ensuring 
the Standby Liquid Control system is capable of mitigating the 
consequences of an ATWS event.
    2. Do the proposed changes create the possibility of a new or 
different type of accident from any previously evaluated?
    The proposed changes to the Unit 1 and Unit 2 Technical 
Specifications do not create the possibility of a new or different 
type of accident from any previously evaluated. The changes are 
administrative in nature only and are intended to clarify Condition 
A of LCO 3.1.7. They ensure the proper condition is entered when 
expected and the sodium pentaborate solution temperature, 
concentration, and volume limits are not exceeded without 
appropriate actions being taken. Those limits, the conditions under 
which the Standby Liquid Control system is required to be operable, 
and the operation of the system remain unchanged and will continue 
to be as described, assumed, and analyzed in the Unit 1 and Unit 2 
Final Safety Analysis Reports, sections 3.8 and 4.2.3.4, 
respectively. The only result of the proposed changes is to reduce 
the time limit for continued unit operation with sodium pentaborate 
solution temperature, concentration, or volume outside Region A and 
Region B from 72 hours to eight hours. Consequently, the possibility 
of a new or different type of accident can not be created by these 
changes.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed changes to the Unit 1 and Unit 2 Technical 
Specifications do not involve a reduction in the margin of safety. 
The changes are administrative in nature only and are intended to 
clarify Condition A of LCO 3.1.7. They ensure the proper condition 
is entered when expected and the sodium pentaborate solution 
temperature, concentration, and volume limits are not exceeded 
without appropriate actions being taken. Those limits, the 
conditions under which the Standby Liquid Control system is required 
to be operable, and the operation of the system remain unchanged by 
the proposed changes and will continue to be as described, assumed, 
and analyzed in the Unit 1 and Unit 2 Final Safety Analysis Reports. 
Therefore, the margin of safety, that is, the ability to bring the 
reactor to a subcritical condition under its most reactive 
conditions with the Standby Liquid Control system, as embodied by 
the sodium pentaborate solution temperature, concentration, and 
volume limits and the system operability requirements will not be 
reduced.
    In conclusion, this proposed license amendment involves no 
significant hazards consideration as determined by the standards set 
forth by the NRC in 10 CFR 50.92(c). Specifically, it has been shown 
in the preceding paragraphs that the proposed changes:
    1. Do not involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. Do not create the possibility of a new or different type of 
accident from any previously evaluated, and

[[Page 46450]]

    3. Do not involve a significant reduction in the margin of 
safety.

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

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of amendment request: July 28, 1999.
    Description of amendment request: The proposed amendment would add 
to the Technical Specifications (TS), new limiting conditions for 
operation and surveillance requirements for the Oscillation Power Range 
Monitor (OPRM) instrumentation installed in response to Generic Letter 
94-02.
    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:

    TVA has concluded that operation of BFN [Brown Ferry Nuclear 
Plant] Unit 3 in accordance with the proposed change to the TS does 
not involve a significant hazards consideration. TVA's conclusion is 
based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of 
the three standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment is to enable the OPRM Upscale trip 
function which is contained in the previously installed PRNM [Power 
Range Neutron Monitoring] equipment. Enabling the OPRM hardware 
provides the long term stability solution required by Generic Letter 
94-02. This hardware incorporates the Option III detect and suppress 
solution reviewed and approved by the NRC in NEDO-31960, ``BWROG 
Long Term Stability Solutions Licensing Methodology.'' The OPRM is 
designed to meet all requirements of GDC 10 and 12 by automatically 
detecting and suppressing design basis thermal-hydraulic power 
oscillations prior to violating the fuel MCPR [minimum critical 
power ratio] Safety Limit. The OPRM system provides this protection 
in the region of the power-to-flow map where instabilities can 
occur, including the region where ICAs [Interim Corrective Actions] 
previously restricted operation because of stability concerns. Thus, 
the ICA restrictions on plant operations are deleted from the TS, 
including region avoidance and the requirement for the operator to 
manually scram the reactor with no recirculation loops operating. 
Operation at high core powers with low core flows may cause a 
slight, but not significant, increase in the probability that an 
instability can occur. This slight increase is acceptable because 
subsequent to the automatic detection of a design basis instability, 
the OPRM Upscale trip provides an automatic scram signal to the RPS 
which is faster protection than the operator initiated manual scram 
required by the current ICAs. Because of this rapid automatic 
action, the consequences of an instability event are not increased 
as a result of the installation of the OPRM system because it 
eliminates operator actions.
    Based on the above discussion, the proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment permits BFN to enable the OPRM power 
oscillation detect and suppress function provided in previously 
installed PRNM hardware, and it simultaneously deletes certain 
restrictions which preclude operation in regions of the power-to-
flow map where oscillations potentially may occur. Enabling the OPRM 
Upscale trip function does not create any new system hardware 
interfaces nor create any new system interactions. Potential 
failures of the OPRM Upscale trip result either in failure to 
perform a mitigation action or in spurious initiation of a reactor 
scram. These failures would not create the possibility of a new or 
different kind of accident. Based on the above discussion, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The OPRM Upscale trip function implements BWROG Stability Option 
III, which was developed to meet the requirements of GDC 10 and GDC 
12 by providing a hardware system that detects the presence of 
thermal-hydraulic instabilities and automatically initiates the 
necessary actions to suppress the oscillations prior to violating 
the MCPR Safety Limit. The NRC has reviewed and accepted the Option 
III methodology described in Licensing Topical Report NEDO-31960 and 
concluded this solution will provide the intended protection. 
Therefore, it is concluded that there will be no reduction in the 
margin of safety as defined in TS as a result of enabling the OPRM 
Upscale trip function and simultaneously removing the operating 
restrictions previously imposed by the ICAs.
    Based on the above discussion, the proposed amendment 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.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: July 20, 1999.
    Description of amendment request: The licensee proposed the 
following five changes: (1) Figure 2.1-1, average power range monitor 
(APRM) Flow Reference Scram and APRM Rod Block Settings, the clarifying 
statement ``Setpoints shall be [less than or equal to] values shown on 
the graph'' is proposed to be added; (2) Bases Section 2.1.B, page 16, 
and Bases Section 3.2 APRM rod block trip discussion, page 77, the 
current Bases is proposed to be replaced with a more accurate 
discussion of the function, as identified in the Vermont Yankee Nuclear 
Power Station (VY) Final Safety Analysis Report (FSAR); (3) Table 
3.1.1, Reactor Protection System (Scram) Instrument Requirements, APRM 
Upscale (Flow Bias) function, it is proposed to add ``with a maximum of 
120%'' to the APRM High Flux (Flow Bias) Trip Function equation; (4) 
For Table 3.2.5, Control Rod-Block Instrumentation, Rod-Block Monitor 
(RBM) Upscale (Flow Bias) function, the caveat ``with a maximum as 
defined in the COLR'' [Core Operating Limits Report] is added to the 
Trip Setting equation; (5) For Bases page 77, it is proposed to delete 
the current paragraph describing the control rod-block systems and 
replace it with the following: ``The trip logic for the nuclear 
instrumentation control rod block logic is 1 out of n; i.e., any trip 
on one of the six APRMs, six IRMs [intermediate range monitors] or four 
SRMs [source range monitors] will result in a rod block. The minimum 
instrument channel requirements for the IRM may be reduced by one for a 
short period of time to allow for maintenance, testing, or calibration. 
The RBM is

[[Page 46451]]

credited in the Continuous Rod Withdrawal During Power Range Operation 
transient for preventing excessive control rod withdrawal before the 
fuel cladding integrity safety limit [minimum critical power ratio] 
(MCPR) or the fuel rod mechanical overpower limits are exceeded. The 
RBM upper limit is clamped to provide protection at greater than 100% 
rated core flow. The clamped value is cycle specific; therefore, it is 
located in the Core Operating Limits Report.''
    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.
    Changes 1 and 3 are administrative and have no impact on 
technical content; therefore, they do not increase the probability 
or consequences of an accident previously evaluated.
    Changes 2 and 5 clarify ambiguities in the Bases. The wording is 
descriptive only and does not change the meaning or intent of the 
specification. Therefore, these changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Change 4 adds the Rod Block Monitor Upscale (Flow Bias) maximum 
value limitation to the Technical Specifications. Limiting the 
upscale trip setting at flows in excess of 100% of rated core flow 
ensures the assumptions of the Continuous Rod Withdrawal During 
Power Range Operation Transient are met. No other accident or 
transient analyses are affected. Therefore, this 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.
    Change 4, limiting the maximum value for the Rod Block Monitor 
Upscale (Flow Bias) function, is a change to plant design, in that 
it clamps the upscale trip setting at flows in excess of 100% of 
rated core flow at the 100% core flow value. This change ensures the 
assumptions of the Continuous Rod Withdrawal During Power Range 
Operation Transient are met and has no effect on any other accident 
or transient analyses. Changes 1, 2, 3, and 5 do not involve a 
change to the plant design.
    None of the proposed changes affects 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, other than the Rod Block Monitor as discussed 
above, are altered as a result of these changes.
    Based on the above VY has concluded that the proposed change 
will not create the possibility of a new or different kind of 
accident from those previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    Changes 1 and 3 are administrative and have no impact on 
technical content. Therefore, they have no effect on margin of 
safety.
    Changes 2 and 5 clarify ambiguities in the Bases, using wording 
taken directly from the FSAR. The wording is descriptive only and 
does not change the meaning or intent of the specification. 
Therefore, these changes do not involve a significant reduction in a 
margin of safety.
    Change 4 adds the Rod Block Monitor Upscale (Flow Bias) maximum 
value limitation to the Technical Specifications. Limiting the 
upscale trip setting at flows in excess of 100% of rated core flow 
ensures the assumptions and, therefore the margin of safety, of the 
Continuous Rod Withdrawal During Power Range Operation transient are 
met. No other accident or transient analyses are affected. 
Therefore, this 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.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    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 notice was 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.

Carolina Power & Light Company, et al., Docket No. 50-261, H. B. 
Robinson Steam Electric Plant, Unit 2, Darlington County, South 
Carolina

    Date of amendment request: July 30, 1999.
    Brief Description of amendment: The proposed amendment would revise 
Required Action A.1 of Technical Specification Limiting Condition for 
Operation 3.7.8, ``Ultimate Heat Sink (UHS),'' to allow a Completion 
Time of 72 hours to restore service water temperature to less than or 
equal to 95 deg.F prior to entering the required actions for plant 
shutdown. The amendment request was proposed as a temporary change to 
be in effect until September 30, 1999.
    Date of publication of individual notice in the Federal Register: 
August 10, 1999 (64 FR 43406).
    Expiration date of individual notice: August 24, 1999, for 
comments; September 8, 1999, for hearings.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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

[[Page 46452]]

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 at the local public document rooms for 
the particular facilities involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of application for amendments: May 23, 1997, as supplemented 
September 27, 1998, and May 26, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications to allow the installation of ABB Combustion 
Engineering leak tight sleeves in defective steam generator tubes as a 
tube repair method.
    Date of issuance: August 5, 1999.
    Effective date: August 5, 1999, to be implemented within 45 days.
    Amendment Nos.: Unit 1--120, Unit 2--120, Unit 3--120.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32285).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 5, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendments: March 14, 1997.
    Brief description of amendments: The amendments deleted license 
conditions which have been satisfied, revise others to delete parts 
which are no longer applicable or to revise references, and make 
editorial changes.
    Date of issuance: August 10, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 110.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendments 
revised the Licenses.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19966).
    The Commission's related evaluation of the amendments is contained 
in an Environmental Assessment dated July 7, 1999 (64FR36722), and a 
Safety Evaluation dated August 10, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.

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

    Date of application for amendments: March 30, 1999, as supplemented 
June 30, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications, Section 3/4.6.G, ``Leakage Detection 
Systems,'' to allow an alternate methodology for quantifying Reactor 
Coolant System (RCS) leakage when the normal RCS leakage detection 
system is inoperable.
    Date of issuance: August 4, 1999.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 189 & 186.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24194).
    The June 30, 1999, submittal provided additional clarifying 
information that did not change the original proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 4, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

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

    Date of application for amendments: March 3, 1999, as supplemented 
May 27, and June 22, 1999.
    Brief description of amendments: The amendments change the required 
qualifications for operations management specified in the technical 
specifications (TSs) for the Beaver Valley Power Station, Units 1 and 2 
(BVPS-1 and BVPS-2). The requirement that the operations manager hold a 
Senior Reactor Operator (SRO) license at the time of appointment is 
changed in the TSs to require that the assistant operations managers, 
one for each unit, hold an SRO license on their assigned unit. The 
revised TSs require the operations manager to hold, or have held, an 
SRO license on a pressurized water reactor. Additionally, the Updated 
Final Safety Analysis Report (UFSAR) for each unit is changed to 
require the operations manager to ``hold, or have held,'' an SRO 
license rather than ``hold'' a license. The revised UFSARs require the 
same as the TSs; that the assistant operations managers hold an SRO 
license on the unit to which they are assigned. Finally, the amendments 
substitute generic personnel titles for plant-specific personnel titles 
in the BVPS-1 and BVPS-2 TSs. The correlation between generic titles 
and plant-specific titles is provided in the revised BVPS-2 UFSAR.
    Date of issuance: August 10, 1999.
    Effective date: Both units, as of date of issuance and shall be 
implemented within 60 days.
    Amendment Nos.: 224 and 100.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19556).
    The May 27, and June 22, 1999, letters provided additional 
information but did not change the initial proposed no significant 
hazards consideration determination or expand the amendment beyond the 
scope of the initial notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 10, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit 3, Citrus County, Florida

    Date of application for amendment: November 24, 1998, as 
supplemented June 23, 1999.
    Brief description of amendment: The amendment approves the addition 
of a safety-related diesel-driven emergency feedwater pump (EFP-3) as a 
functional replacement for the existing motor-driven pump, addition of 
technical specifications and surveillances for this new pump, and 
deletion of cycle specific interim technical specifications which would 
not be required after the addition of the new pump.

[[Page 46453]]

    Date of issuance: August 11, 1999.
    Effective date: As of the date of issuance and shall be implemented 
prior to commencing cycle 12 operation.
    Amendment No.: 182.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 13, 1999 (64 FR 
2247).
    The supplemental letter dated June 23, 1999, did not change the 
original proposed no significant hazards consideration determination, 
or expand the scope of the amendment request as originally noticed. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 11, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

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

    Date of amendment request: November 4, 1998.
    Description of amendment request: To revise Technical 
Specifications Surveillance Requirement 4.5.2b.1 to delete the 
prescribed method of venting the Emergency Core Cooling System (ECCS) 
which would allow an alternate method to verify that the ECCS piping is 
full of water. In addition, the associated Bases are being revised to 
reflect the intent of the surveillance requirement.
    Date of issuance: August 12, 1999.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 61.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4157)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 12, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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

    Date of application for amendment: June 4, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications by extending the allowed outage time for the 32 
emergency diesel generator and its fuel oil storage tank.
    Date of issuance: August 9, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 190.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1999 (64 FR 
36408).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: January 25, 1996, as 
supplemented April 26, 1996, September 12, 1996, March 17, 1997, 
September 9, 1997, December 30, 1998, and May 19, 1999.
    Brief description of amendment: The amendment extends the allowed 
outage time for an emergency diesel generator (EDG) system from 7 to 14 
days, revises requirements for EDG testing at power, and revises 
electrical power requirements for cold shutdown and refueling modes.
    Date of issuance: July 30, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 253.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notices in Federal Register: March 27, 1996 (61 FR 
13532) and June 30, 1999 (64 FR 35208).
    The licensee provided additional information on April 26, 1996, 
September 12, 1996, March 17, 1997, September 9, 1997, and December 30, 
1998, that provided clarifying information within the scope of the 
initial Federal Register notice and did not change the staff's original 
proposed no significant hazards consideration determination. The 
changes proposed on May 19, 1999, were reflected in the staff's revised 
proposed finding of no significant hazards consideration, and encompass 
the additional information provided by the licensee.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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: June 19, 1998, (Unit 1) and 
August 5, 1998, (Unit 2) as supplemented by letter dated November 23, 
1998.
    Brief description of amendments: The amendments to the Unit 1 and 
Unit 2 Technical Specifications (TSs) involve the addition of a new 
section entitled ``Oscillation Power Range Monitoring (OPRM) 
Instrumentation'' and revisions to Section 3.4.1 ``Recirculation Loops 
Operating'' to remove the specifications related to thermal power 
stability which are no longer required after the installation of OPRM 
instrumentation.
    Date of issuance: July 30, 1999.
    Effective date: Effective as of its date of issuance and is to be 
implemented within 90 days following startup from the Unit 2 ninth 
Refueling Inspection Outage, currently scheduled for April 16, 1999.
    Amendment Nos.: 184 and 188.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43210) and August 26, 1998 (63 FR 45528).
    The November 23, 1998, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the amendment request beyond the 
scope of the initial notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 30, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

[[Page 46454]]

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

    Date of application for amendments: April 23, 1999.
    Brief description of amendments: The amendment changes Permanently 
Defueled Technical Specification 
D3/4.1, ``Spent Fuel Pool Level,'' to replace a specific reference to 
spent fuel pool (SFP) level alarm switches with a generic reference to 
SFP level instrumentation.
    Date of issuance: August 13, 1999.
    Effective date: August 13, 1999, to be implemented within 30 days.
    Amendment No.: 126.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35210).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Central Library, Government 
Documents, 828 I Street, Sacramento, California 95814.

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

    Date of amendment request: March 2, 1999, as supplemented by letter 
dated July 13, 1999.
    Brief description of amendments: The amendments allow the use of a 
``check valve with flow through the valve secured'' as an additional 
means to isolate an affected containment penetration (i.e., a 
penetration with an inoperable penetration barrier) in Technical 
Specification 3.6.3, Action b.
    Date of issuance: August 3, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-113; Unit 2-101.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17030).
    The July 13, 1999, supplement provided additional clarifying 
information within the scope of the original notice and did not change 
the staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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 12, 1997, as supplemented 
by letters dated March 30, 1999, April 23, 1999, and June 18, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to extend, from 7 days to 14 days, the 
Allowable Outage Time applicable to an inoperable emergency diesel 
generator.
    Date of issuance: August 2, 1999.
    Effective date: August 2, 1999.
    Amendment Nos.: 259 and 218.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the TS.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35211)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 2, 1999.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.

TU Electric Company, 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 4, 1999, as supplemented by letter 
dated June 4, 1999.
    Brief description of amendments: The amendments correct a number of 
editorial errors in the Technical Specifications that occurred with the 
issuance of Amendment No. 64 to Facility Operating License Nos. NPF-87 
and NPF-89, regarding the improved Technical Specifications conversion. 
In addition, Surveillance Requirement (SR) 3.8.4.7 is revised to allow 
the substitution of a modified performance discharge test, for a 
service test, for the 125 VDC batteries and SRs 3.8.1.7, 3.8.1.12, 
3.8.1.15, and 3.8.1.20 are revised to separate the voltage and 
frequency acceptance criteria for the diesel generator start 
surveillances into two sets of criteria; those criteria required to be 
met within 10 seconds, and those criteria required to be met following 
achievement of steady state conditions.
    Date of issuance: August 3, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-Amendment No. 66; Unit 2-Amendment No. 66.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29715); and June 30, 1999 (64 FR 35212).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of application for amendment: February 1, 1999, as 
supplemented on April 19 and April 23, 1999.
    Brief description of amendment: The amendment totally replaces the 
current Technical Specifications Section 6.0, ``Administrative 
Controls.'' Administrative changes to certain other sections of the 
Technical Specifications were made to conform to the changes resulting 
from the re-write of Section 6.0.
    The changes represent a comprehensive upgrade of Section 6.0 of the 
Vermont Yankee Technical Specifications, incorporating improvements in 
content and format based on industry standards. In accordance with 
industry practice, some Technical Specifications requirements are being 
relocated to the recently implemented Vermont Yankee Technical 
Requirements Manual, Offsite Dose Calculation Manual, or Vermont Yankee 
Operational Quality Assurance Manual and are being eliminated from the 
Technical Specification.
    Date of Issuance: July 19, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 171.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27326).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 19, 1999.
    No significant hazards consideration comments received: No.

[[Page 46455]]

    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: June 3, 1999, as supplemented by 
letter dated July 22, 1999.
    Brief description of amendment: The amendment updates the operating 
license to reflect the name change of the licensee from ``Washington 
Public Power Supply System'' to ``Energy Northwest'' and the name 
change of the facility from ``WPPSS Nuclear Project No. 2'' to ``WNP-
2.''
    Date of issuance: August 2, 1999.
    Effective date: August 2, 1999.
    Amendment No.: 157.
    Facility Operating License No. NPF-21: The amendment revised the 
operating license.
    Date of initial notice in Federal Register: June 30, 1999. (64 FR 
35214).
    The July 22, 1999, supplemental letter provided additional 
clarifying information, did not significantly 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 August 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

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

    Date of application for amendments: January 29, 1999. (TSCR 211), 
as supplemented June 9 and July 15, 1999.
    Brief description of amendments: These amendments reflect changes 
to Sections 15.6 and 15.7 of the Point Beach Nuclear Plant, Units 1 and 
2, Technical Specifications (TSs). The changes are considered 
administrative in nature and reflect personnel title changes, an 
increase in minimum operating crew shift staffing, relocation of the 
Manager's Supervisory Staff composition and functional requirements to 
owner-controlled documents, and revisions to the procedure review and 
approval process.
    Date of issuance: August 11, 1999.
    Effective date: August 11, 1999. The TSs shall be implemented 
within 90 days. Implementation also includes removal of selected 
requirements from TS Section 15.6, Administrative Controls, and the 
relocation of other requirements to licensee-controlled documents as 
described in the licensee's application dated January 29, 1999, as 
supplemented June 9 and July 15, 1999, and evaluated in the staff's 
safety evaluation attached to the amendments. With respect to changes 
to the final safety analysis report (FSAR), Wisconsin Electric Power 
Company shall incorporate the revisions into the next FSAR update in 
accordance with the schedule in 10 CFR 50.71(e).
    Amendment Nos.: Unit 1-190; Unit 2-195.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9202).
    The June 9 and July 15, 1999, letters provided additional 
clarifying information within the scope of the original Federal 
Register notice and did not affect the staff's initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 11, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

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

    Date of amendment request: June 10, 1999.
    Brief description of amendment: The amendment revised Technical 
Specification Table 3.3-4, Functional Unit 7.b., Automatic Switchover 
to Containment Sump (Refueling Water Storage Tank Level--Low-Low) to 
reflect the results of calculations that were performed for the 
associated instrumentation setpoints to consider the density variations 
due to temperature and boric acid concentrations.
    Date of issuance: August 9, 1999.
    Effective date: August 9, 1999, and shall be implemented within 60 
days from issuance of the amendment.
    Amendment No.: 126.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35215).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

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

    Date of amendment request: June 11, 1999.
    Brief description of amendment: The amendment revises TS 3.7.1.6, 
Steam Generator Atmospheric Relief Valves, and associated Bases to (1) 
require four atmospheric relief valves (ARVs) to be operable, (2) 
eliminate the use of ``required'' in the action statements, (3) provide 
action statements to address inoperability of two ARVs and three or 
more ARVs due to causes other than excessive leakage, and (4) limit the 
Limiting Condition for Operation 3.0.4 exception to when one ARV is 
inoperable due to causes other than excessive seat leakage.
    Date of issuance: August 12, 1999.
    Effective date: August 12, 1999, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 127.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35215).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

    Dated at Rockville, Maryland, this 18th day of August 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-21914 Filed 8-24-99; 8:45 am]
BILLING CODE 7590-01-P