[Federal Register Volume 65, Number 36 (Wednesday, February 23, 2000)]
[Notices]
[Pages 9000-9017]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-4236]


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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 January 29, 2000, through February 11, 2000. 
The last biweekly notice was published on February 9, 2000 (65 FR 
6402).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not: 
(1) Involve a significant increase in the probability or consequences 
of an accident previously evaluated; or (2) create the possibility of a 
new or different kind of accident from any accident previously 
evaluated; or (3) involve a significant reduction in a margin of 
safety. The basis for this proposed determination for each amendment 
request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By March 24, 2000, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be

[[Page 9001]]

affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.714 which 
is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of amendment request: May 27, 1999.
    Description of amendment request: The requested amendment proposes 
to increase the maximum allowable Service Water (SW) temperature used 
to determine operability of the Ultimate Heat Sink (UHS) from 95  deg.F 
to 97  deg.F. The amendment includes all the TS changes necessary as a 
result of new analyses performed to support the increase of the maximum 
allowable SW temperature.
    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:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
Technical Specification change and has concluded that it does not 
involve a significant hazards consideration. The conclusion is in 
accordance with the criteria set forth in 10 CFR 50.92. The bases 
for the conclusion that the proposed change does not involve a 
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change increases the maximum allowable Service 
Water (SW) temperature, which is used to determine OPERABILITY of 
the Ultimate Heat Sink (UHS), from 95  deg.F to 97  deg.F. As a 
result of the new analyses to support the increase in SW 
temperature, the proposed change also decreases the required 
actuation setpoint for the Containment Pressure High High signal 
from 20 psig to 10 psig, decreases the closure time credited for the 
Main Feedwater

[[Page 9002]]

Isolation Valves (MFIVs) in the analysis from 80 seconds to 50 
seconds, increases the required operating pressure for the Isolation 
Valve Seal Water (IVSW) and IVSW nitrogen bottle pressure from 44 
psig to 44.6 psig, decreases the closure time for Main Steam 
Isolation Valves (MSIVs) credited in the analysis from 5 seconds to 
2 seconds, and increases the peak calculated containment internal 
pressure for a large break Loss of Coolant Accident (LOCA), Pa, from 
40 psig to 40.5 psig. In addition, the Containment Spray (CS) 
actuation circuitry will be modified to allow the CS pumps to be 
restarted after they have been stopped while the original actuation 
signal is present.
    SW temperature is not itself an initiator of accidents evaluated 
in the Safety Analysis report (SAR). The components provided SW flow 
that are required to perform a safety-related function are designed 
to operate at temperatures above the temperatures to which SW will 
be increased. Therefore, these components are not more likely to 
fail and initiate an accident. The components have been shown to 
perform their intended safety related function with the higher SW 
temperatures. Containment analyses have been performed that show 
that containment integrity and equipment environmental qualification 
are maintained.
    The modification to the Containment High High Pressure actuation 
setpoint will not increase the probability of an unwanted actuation. 
Changing the actuation setpoint will not change the reliability of 
this function. The Containment Pressure High High Pressure function 
will (1) initiate Containment Spray sooner, which will mitigate the 
pressure and temperature transient sooner, and (2) isolate leakage 
of radioactivity from containment through ``essential'' process 
lines sooner in an accident. Also, the lower actuation setpoint, in 
conjunction with other analysis assumptions, has been evaluated to 
result in a slight decrease (-2  deg.F) in the large break LOCA Peak 
Cladding Temperature.
    Crediting faster MFIV closure in the Main Steam Line Break 
(MSLB) containment analysis will not change the probability of MFIV 
failure or the probability that the MFIV will initiate an accident 
because a physical modification is not associated with the proposed 
change. (The physical modification is being implemented in 
accordance with 10 CFR 50.59). Since there is no physical 
modification, the amount of feedwater addition to containment during 
[an] MSLB if the Main Feedwater Regulating Valve (MFRV) fails [to] 
open will not change, although the amount calculated by the analysis 
will be reduced.
    Crediting faster MSIV closure in the MSLB containment analysis 
will not change the probability of MSIV failure or the probability 
that the MSIV will initiate an accident because a physical 
modification is not involved. Since there is no physical 
modification, the amount of blowdown from the unaffected SGs [steam 
generators] and the amount of radioactivity released to the 
environment by [an] MSLB will not be adversely affected, although 
the amount calculated by the analysis will be reduced. Crediting a 
faster closure time does not require crediting a faster MSIV opening 
time because of the valve design, and opening [an] MSIV is not 
postulated for an analyzed accident.
    Changing the minimum operating pressure of the IVSW components 
does not involve a physical modification, hence, will not affect the 
probability that components will fail or initiate an accident. The 
IVSW system will perform its containment isolation function by 
providing a water seal at the higher pressure calculated by the new 
large break LOCA containment analysis.
    The Containment Leakage Rate Testing (CLRT) program historically 
has performed integrated leak rate testing and local leak rate 
testing at pressures higher than the peak containment pressure 
calculated by the new large break LOCA containment analysis. The 
components which are tested by the CLRT program are designed for 
operation at a pressure higher than the pressure to which they are 
tested. The current CLRT program ensures that the containment 
leakage is less than that used to calculate the doses for a large 
break LOCA accident.
    The modification to the CS actuation circuitry will not affect 
the reliability of the circuit. The modification will be tested 
periodically to ensure reliability and to confirm the capability of 
restoring CS after being blocked. Blocking the actuation circuitry 
will be procedurally controlled and will allow the CS pumps to be 
restarted, after being stopped, when an actuation signal is present. 
The analysis results show that containment pressure and temperature 
are within design limits when CS is stopped for the switchover.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated in the SAR.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The components provided SW flow have been shown to perform their 
safety related function with the higher service temperature, hence, 
will not exhibit any new type of failure mechanism or mode as a 
result of the increased temperatures.
    Decreasing the Containment High High Pressure actuation setpoint 
only changes the time at which the signal is generated, not how it 
is generated, or how the actuated equipment responds to the signal, 
hence, will not introduce any new types of failures.
    Crediting faster MFIV and MSIV stroke times in the MSLB 
containment analysis does not involve a physical modification, 
hence, can not introduce any new failure modes.
    The IVSW components and the components tested by the CLRT 
program are designed for pressures that are higher than the 
pressures at which they are proposed to operate and be tested. As 
the functions of these components are not changing, and the 
components are capable of withstanding the higher pressure, a higher 
operating or testing pressure will not create any new failure 
mechanisms or accidents.
    The modification to the CS actuation circuitry will be tested 
periodically to ensure proper operation and reliability of the 
circuit. Even if one of the blocking circuits should fail during 
operation, a single failure of a CS pump has been considered in the 
containment analysis, hence, is not a new type of failure or 
accident.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Containment structural integrity, containment leakage, fuel 
cladding, equipment environmental qualification, EDG electrical 
capacity, and UHS capability were considered to determine if the 
proposed change involves a significant reduction in a margin of 
safety.
    Containment pressure is limited to the design pressure of 42 
psig to maintain structural integrity. A structural integrity test 
at 115% of the design pressure (48.3 psig) has confirmed the 
containment's structural capability. The new containment analyses 
for large break LOCA and MSLB using [an] SW temperature of 100 
deg.F show that the containment pressure does not exceed 42 psig. 
The margin of safety for containment is not reduced by the proposed 
change because the design pressure is not exceeded. The containment 
leakage rate, La, is limited to 0.1% of the containment air weight 
per day. La is based on the peak calculated containment internal 
pressure, Pa, for the design basis LOCA. The offsite doses resulting 
from an accident are based on La. If containment leakage does not 
exceed La, the margin of safety is not reduced. The leakage rates 
for Type A, B, and C containment penetrations are measured 
periodically throughout plant life to ensure that containment 
leakage is [less than or equal to] La. The leakage rate acceptance 
criteria are [less than or equal to] 0.75 L for Type A tests, and 
[less than or equal to] 0.60 La for Type B and Type C tests. As a 
result of using [an] SW temperature of 97  deg.F in the new large 
break LOCA containment analysis, Pa has changed from 40 psig to 40.5 
which changes the pressure at which the Type A, B, and C containment 
penetration leakage is measured. Historically, containment leakage 
rate testing has been performed at the containment design pressure 
of 42 psig or higher. The margin of safety related to containment 
leakage is not reduced by the proposed change because containment 
leakage is [less than or equal to] La.
    Fuel cladding integrity is evaluated by determining the effect 
on the Peak Cladding Temperature (PCT) and the Departure to Nucleate 
Boiling Ratio (DNBR) for postulated accident. The PCT for a large 
break LOCA changes by -2  deg.F as a result of the proposed change 
including associated changes. The DNBR for a non-limiting case of 
the MSLB changes, but the margin to the DNBR limit is very large. 
Therefore, fuel cladding integrity is not adversely affected.
    Safety-related equipment is potentially required to function in 
an adverse environment during and following an accident. Using [an] 
SW temperature of 97  deg.F, the new large break LOCA and MSLB 
containment analyses yield temperature and pressure profiles show 
that the temperature and pressure profiles for equipment required to 
operate during and following an accident

[[Page 9003]]

are qualified. The margin of safety related to equipment 
environmental qualification is not reduced by the proposed change 
because equipment required to operate during and following an 
accident are environmentally qualified.
    The Emergency Diesel Generators (EDGs) provide emergency 
electrical power to run safety-related equipment following an 
accident that is accompanied by a loss of offsite power. The EDGs 
are rated at 110% capacity for 2 hours out of each 24 hours and 
tested between 106% to 110% for at least 1.75 hours. Since the EDG 
can provide 110% for 1.75 hours, the margin of safety is not 
reduced. Using [an] SW temperature of 97  deg.F, a calculation shows 
that adequate cooling is provided for the EDG to produce 110% 
electrical output.
    The UHS is required to provide cooling water for at least 22 
days following a design basis accident. The UHS is able to provide 
cooling water for 22.1 days at a temperature of 100  deg.F. 
Therefore, the cooling capability of the UHS would not be adversely 
affected.
    Based on the above, it may be concluded that the proposed change 
does not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of amendment request: December 22, 1999.
    Description of amendment request: The proposed amendment would 
expand the Core Operating Limits Report (COLR) and relocate reactor 
coolant system related cycle-specific parameter limits from the 
technical specifications (TSs) and include them in the COLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes are programmatic and administrative in 
nature which do not physically alter safety-related systems, nor 
affect the way in which safety-related systems perform their 
functions. The proposed changes remove cycle-specific parameter 
limits from TS 3.4.1 and relocate them to the COLRs which do not 
change plant design or affect system operating parameters. In 
addition, the minimum limit for [Reactor Coolant System] RCS total 
flow rate is being retained in TS 3.4.1 to assure that a lower flow 
rate than reviewed by the NRC will not be used. The proposed changes 
do not, by themselves, alter any of the parameter limits. The 
removal of the cycle-specific parameter limits from the TS does not 
eliminate existing requirements to comply with the parameter limits. 
The existing TS Section 5.6.5b, COLR Reporting Requirements, 
continues to ensure that the analytical methods used to determine 
the core operating limits meet NRC reviewed and approved 
methodologies. The existing TS Section 5.6.5c, COLR Reporting 
Requirements, continues to ensure that applicable limits of the 
safety analyses are met. Further, more specific requirements 
regarding the safety limits (i.e., [Departure from Nucleate Boiling 
Ratio] DNBR limit and peak fuel centerline temperature limit) are 
being imposed in TS 2.1.1, ``Reactor Core Safety Limits,'' replacing 
the Reactor Core Safety Limits (RCSL) figure which are consistent 
with the values stated in the Updated Final Safety Analysis Report 
(UFSAR).
    Although the relocation of the cycle-specific parameter limits 
to the COLRs would allow revision of the affected parameter limits 
without prior NRC approval, there is no significant effect on the 
probability or consequences of an accident previously evaluated. 
Future changes to the COLR parameter limits could result in event 
consequences which are either slightly less or slightly more severe 
than the consequences for the same event using the present parameter 
limits. The differences would not be significant and would be 
bounded by the existing requirement of TS Section 5.6.5c to meet the 
applicable limits of the safety analyses.
    The cycle-specific parameter limits being transferred from the 
TS to the COLRs will continue to be controlled under existing 
programs and procedures. The UFSAR accident analyses will continue 
to be examined with respect to changes in the cycle-dependent 
parameters obtained using NRC reviewed and approved reload design 
methodologies, ensuring that the transient evaluation of new reload 
designs are bounded by previously accepted analyses. This 
examination will continue to be performed pursuant to 10 CFR 50.59 
requirements ensuring that future reload designs will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Additionally, the proposed changes do 
not allow for an increase in plant power levels, do not increase the 
production, nor alter the flow path or method of disposal of 
radioactive waste or byproducts. Therefore, the proposed changes do 
not change the types or increase the amounts of any effluents 
released offsite.
    Therefore, the proposed changes do not involve an increase in 
the probability or consequences of an 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 changes that retain the minimum limit for RCS total 
flow rate in the TS, and that relocate certain cycle-specific 
parameter limits from the TS to the COLR, thus removing the 
requirement for prior NRC approval of revisions to those parameters, 
do not involve a physical change to the plant. No new equipment is 
being introduced, and installed equipment is not being operated in a 
new or different manner. There is no change being made to the 
parameters within which the plant is operated, other than their 
relocation to the COLRs. There are no setpoints affected by the 
proposed changes at which protective or mitigative actions are 
initiated. The proposed changes will not alter the manner in which 
equipment operation is initiated, nor will the function demands on 
credited equipment be changed. No alteration in the procedures which 
ensure the plant remains within analyzed limits is being proposed, 
and no change is being made to the procedures relied upon to respond 
to an off-normal event. As such, no new failure modes are being 
introduced.
    Relocation of cycle-specific parameter limits has no influence 
or impact on, nor does it contribute in any way to the possibility 
of a new or different kind of accident. The relocated cycle-specific 
parameter limits will continue to be calculated using the NRC 
reviewed and approved methodology. The proposed changes do not alter 
assumptions made in the safety analysis and operation within the 
core operating limits will continue.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes do not physically alter safety-
related systems, nor does it effect the way in which safety-related 
systems perform their functions. The setpoints at which protective 
actions are initiated are not altered by the proposed changes. 
Therefore, sufficient equipment remains available to actuate upon 
demand for the purpose of mitigating an analyzed event. As the 
proposed changes to relocate cycle-specific parameter limits to the 
COLRs will not affect plant design or system operating parameters, 
there is no detrimental impact on any equipment design parameter, 
and the plant will continue to operate within prescribed limits.
    The development of cycle-specific parameter limits for future 
reload designs will continue to conform to NRC reviewed and approved 
methodologies, and will be performed pursuant to 10 CFR 50.59 to 
assure that plant operation within cycle-specific parameter limits 
will not involve a significant reduction in the margin of safety.

[[Page 9004]]

    Therefore, the proposed changes do not involve a reduction in a 
margin of safety.

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

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: September 17, 1999.
    Description of amendment request: The proposed amendment would 
change the Arkansas Nuclear One, Unit 2 (ANO-2) heavy load handling 
requirements and transportation provisions to permit the movement of 
the original and replacement steam generators through the ANO-2 
containment construction opening during the steam generator replacement 
outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    During the 2R14 refueling outage/steam generator replacement 
outage, the OSGs [original steam generators] and the RSGs 
[replacement steam generators] will be moved between the new steam 
generator storage area/original steam generator storage facility and 
the runway beam support system (RBSS)/outside lift system (OLS). The 
RBSS/OLS is the structure used to rig the SGs [steam generators] in 
and out of the reactor containment building. In consideration of the 
magnitude of the loads being handled, the RBSS, OLS and transporters 
are of a robust, rugged design, proven by many prior steam generator 
replacements and other heavy load handling operations. However, due 
to the location of safety related underground structures, systems, 
and components (SCCs) in the vicinity of the RBSS/OLS and along the 
steam generator (SG) haul route, potential load handling accidents 
along the load paths must be considered for their effects on the 
SCCs. At ANO-2, the ground cover over several buried SSCs is not 
sufficient to be able to rule out the potential for a load drop to 
damage or cause failure of these SCCs. The functions of the SSCs in 
question are as support systems to the ANO-1 [Arkansas Nuclear One, 
Unit 1] and ANO-2 emergency diesel generators and the ANO-1 service 
water system. The fire protection system, a non-safety related 
system, was also considered. Existing plant procedures adequately 
address the scenario in question for the fire protection system.
    The cause of a SG drop is assumed to be a non-mechanistic 
failure of the RBSS/OLS (or associated rigging), a failure of the SG 
transporter leveling hydraulics, or a seismically-induced failure of 
the loaded RBSS/OLS or SG transporter. The possibility of drops 
associated with other external events, such as tornadoes, high 
winds, and tornado missiles will be substantially minimized by 
procedures that prevent load handling under these weather 
conditions.
    With ANO-2 defueled, the impact on ANO-2 due to loss of the 
emergency diesel generators fuel oil transfer system will be 
minimal. Long term actions to provide makeup water to the spent fuel 
pool may be necessary, but no immediate actions are required.
    For ANO-1, a steam generator drop could render both diesel 
generators inoperable due to the loss of the fuel oil transfer 
system, and the emergency cooling pond inoperable due to the loss of 
the service water return line to the pond. Since ANO-1 is expected 
to be at full power operation, these conditions would require prompt 
action in accordance with technical specifications. Immediately 
following a drop from the OLS or from the transporter in the 
vicinity of the OLS, where damage to these systems is possible, ANO-
1 will begin a shutdown and cooldown to cold shutdown conditions. In 
conjunction with the unit shutdown, contingency actions to provide 
temporary connections from the fuel oil storage facility to the ANO-
1 emergency diesel generator day tanks, and temporary power to the 
fuel transfer pumps would be implemented.
    The ability of ANO-1 to safely respond to analyzed events would 
be undiminished with the possible exception of the functions 
affected by the damaged equipment. With the compensatory measures to 
be established prior to the steam generator handling operations, and 
with the planned responses to a steam generator drop, the support 
system functions of the diesel generators and the service water 
system can be assumed to be maintained following the drop. 
Therefore, the drop will not affect the consequences of any analyzed 
event.
    While the drop of a steam generator could cause damage to some 
safety related plant equipment, the failures of these components are 
not precursors to any analyzed accident. The drop of a steam 
generator will not have any other impact on plant equipment, and 
thus will not induce any analyzed plant transient. It will, however, 
result in a malfunction of equipment important to safety of a 
different type than any previously evaluated. Based on the 
compensatory measures and the low likelihood of the event during SG 
movement, this temporary condition is considered to be acceptable. 
On these bases, it is concluded that the proposed load handling 
operations will not significantly increase the probability or the 
consequences of accidents previously analyzed.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From Any Previously Evaluated

    As noted in the response to the first question above, the only 
potential for a new or different kind of accident associated with 
this change request arises from a drop of a steam generator which is 
assumed to cause the loss of emergency power support systems for 
ANO-1. The cause of a SG drop is assumed to be a non-mechanistic 
failure of the RBSS/OLS (or associated rigging), a failure of the SG 
transporter leveling hydraulics, or a seismically-induced failure of 
the loaded RBSS/OLS or SG transporter. In the absence of a seismic 
event, there is no initiator for any consequential events (e.g., 
loss of offsite power) other than those directly caused by impact of 
the SG. Given this scenario, the plant response to a SG drop event 
would be governed by the technical specifications and existing plant 
procedures.
    If a SG drop is seismically-induced, the simultaneous loss of 
normal offsite power sources is also assumed in this case since 
these sources are not seismically qualified. While this event is 
very unlikely due to the low frequency of earthquakes and the small 
amount of time that a steam generator will be in a position to cause 
damage, Entergy [Operations, Inc.] will provide contingency plans 
and compensatory measures so that makeup to the ANO-2 spent fuel 
pool and fuel oil supply to the ANO-1 emergency diesel generators 
and transfer pump power supply are assured under any circumstances.
    Availability of the redundant ANO-1 service water heat sink, the 
Dardanelle Reservoir, during a seismic event assures that an 
uninterrupted source of service water will be available to support 
shutdown cooling of ANO-1.
    The proposed load handling plans will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    ANO-1 Technical Specification 3.7.1.C requires both EDGs 
[emergency diesel generators] to be operable when the reactor 
temperature is 200  deg.F. If this condition is not met, 
Limiting Condition for Operation 3.0.3 applies. It requires that 
within one hour, action shall be initiated to place the unit in an 
operating condition in which the specification does not apply by 
placing it, as applicable, in at least hot standby within the next 6 
hours, at least hot shutdown within the following 6 hours, and at 
least cold shutdown within the subsequent 24 hours. The bases for 
technical specification 3.7.1.C indicate that these operability 
requirements ensure that an adequate, reliable power source is 
available for all electrical equipment during startup, normal 
operation, safe shutdown, and handling of all emergency situations. 
The bases for EDG operation also require at least a seven day total 
diesel oil inventory during complete loss of electrical power 
conditions.

[[Page 9005]]

    The postulated loss of both trains of the ANO-1 EDG fuel oil 
transfer system due to a SG drop would require that ANO-1 be shut 
down. This situation could be considered to involve a reduction in 
the margin of safety, because a new common cause failure mechanism 
is being introduced by the movement of the SGs over the EDG fuel oil 
lines and transfer pump power cables. To restore the margin of 
safety and return the EDGs to functionality, temporary compensatory 
measures are being proposed.
    Based on the above discussions, with the implementation of the 
proposed compensatory measures and the low likelihood of such an 
event, the failures caused by a SG drop event will not involve a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: August 18, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) 4.4.5, ``Steam Generators,'' to 
note that the requirements for inservice inspection do not apply during 
the steam generator replacement outage (2R14), to delete inspection 
requirements associated with steam generator tube sleeving and repair 
limits, to extend the inspection interval to a maximum of once per 40 
months provided the inspection results from the first inspection 
following the preservice inspection fall into the C-1 category, to 
revise the preservice inspection requirements on when the hydrostatic 
test and the eddy current inspection of the tubes would be performed, 
and to revise the reporting frequency of the results of steam generator 
tube inspections to within 12 months following completion of the 
inservice inspection. Related changes to the Bases would also be made.
    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:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The accidents of interest are a tube rupture, loss of coolant 
accident (LOCA) in combination with a safe shutdown earthquake and a 
steam line break in combination with a safe shutdown earthquake. A 
reduction in tube integrity could increase the possibility of a tube 
rupture accident and increase the consequences of a steam line break 
or LOCA. The tubing in the replacement steam generators is designed 
and evaluated consistent with the margins of safety specified in the 
ASME [American Society of Mechanical Engineers] Code [Boiler and 
Pressure Vessel Code], Section III. The program for periodic 
inservice inspection provides sufficient time to take proper and 
timely corrective action if tube degradation is present. The ASME 
[Code], Section XI basis for the 40% through wall plugging limit is 
applicable to the replacement steam generators just as it was to the 
original steam generators. As a result there is no reduction in tube 
integrity for the replacement steam generators.
    Addition of a ``Note'' to clarify that inservice inspection is 
not required during the steam generator replacement outage is an 
administrative change that provides clarification regarding 
inservice inspection requirements. The change in reporting 
requirements is also an administrative change. The requirements for 
inservice inspection or the plugging limit for the tubes are not 
altered by these administrative changes. Additionally, changes were 
made to the bases to remove potentially misleading information. 
Bases changes are considered to be administrative in nature.
    Elimination of the repair option and the associated references 
to repair of the original steam generator tubes is an administrative 
adjustment since the sleeve design is not applicable to the 
replacement steam generators. The elimination of the repair option 
does not alter the requirements for inservice inspection or reduce 
the plugging limit for the tubes.
    The proposed change to extend the inspection interval to a 
maximum of once per 40 months is acceptable based on the use of the 
superior Alloy 690 tubing material. Significant industry knowledge 
has been gained from monitoring the performance of steam generators 
that have been replaced. Alloy 690 tubing material has proven to be 
superior to Alloy 600 in regard to corrosion resistance. Plants that 
have utilized Alloy 690 tubing in their replacement steam generators 
have not experienced corrosion-induced degradation.
    A preservice eddy current inspection will be performed onsite 
prior to installation of the replacement steam generators. The 
orientation of the replacement steam generators during the eddy 
current exam will not impact the results. The hydrostatic test 
required by the ASME Code, Section III for the replacement steam 
generators is to be performed in the manufacturing facility and not 
as part of a reactor coolant system hydrostatic test. The post-
repair leakage test required by the ASME Code, Section XI for an 
operating plant is performed at a much lower pressure. No evolutions 
subsequent to the replacement steam generator hydrostatic test are 
expected to occur that will change the condition of the tubes prior 
to operation. This change does not alter the requirement to perform 
a preservice inspection. As a result, an inservice inspection is not 
required during the steam generator replacement outage.
    The requested ANO-2 [Arkansas Nuclear One, Unit 2] Technical 
Specification changes do not alter the requirements for tube 
integrity, tube inspection, or tube plugging limit. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From Any Previously Evaluated

    The proposed changes do not affect the design or function of any 
other safety-related component. There is no mechanism to create a 
new or different kind of accident for the replacement steam 
generators by eliminating repair criteria or by clarifying the 
applicable preservice and inservice inspection requirements because 
a baseline of tube conditions is established and plugging limits are 
maintained to ensure that defective tubes are removed from service. 
A change in inspection frequency has a negligible impact on the pre-
accident state of the reactor core or post accident confinement of 
radionuclides within the containment building. Changing the 
inspection frequency creates no new failure modes or accident 
initiators/precursors.
    The requested ANO-2 Technical Specification changes do not alter 
the requirements for tube integrity, tube inspection or tube 
plugging limit. Therefore, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    The tubing in the replacement steam generators is designed and 
evaluated consistent with the margins of safety specified in the 
ASME Code, Section III. The program for periodic inservice 
inspection provides sufficient time to take proper and timely 
corrective action to preserve the design margin if tube degradation 
is present.
    Due to the superior Alloy 690 tubing material and the 
significant amount of industry knowledge and operating history with 
this improved tubing material, extending the inspection interval to 
a maximum of once per 40 months will still allow the integrity of 
the steam generator tubing to be ensured. The steam generator 
inspection program is not intended to provide an accident mitigation 
or assessment function; therefore, this change results in a neutral 
impact to the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve a significant 
hazards consideration.


[[Page 9006]]


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

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

    Date of amendment request: November 3, 1999.
    Description of amendment request: The proposed amendment would 
increase the containment structural design pressure from 54 to 59 psig, 
revise Technical Specification (TS) Table 3.3-3 to add a containment 
spray actuation signal on high-high containment building pressure to 
terminate main feedwater and main steam flow from the unaffected steam 
generator, revise TS 3.6.1.4 and Figure 3.6-1 to change the allowable 
containment initial conditions to be consistent with analysis 
assumptions, revise TS 4.6.2.1 to increase the allowable containment 
spray pump degradation from 6.3% to 10.0%, and revise TS 6.15 to 
increase the calculated peak accident pressure in the containment 
leakage rate testing program from 54 to 58 psig and to clarify the 
allowable leakage rate. Related changes to the Bases would also be 
made.
    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:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The containment building will meet structural requirements for 
the higher design pressure. Except for the application of CSAS 
[Containment Spray Actuation Signal] in a different manner than used 
previously, the electrical penetration seal modifications and the 
containment cooling fan pitch change, increasing the containment 
structural design pressure is analytical. There are no changes to 
the allowable containment leakage rate. The increase in design 
pressure requires changes to the bases of the technical 
specifications and the SAR [Safety Analysis Report]. However, the 
peak accident and design pressures are below the failure pressure of 
any potentially affected system, structure or component. The change 
does not increase the probability of an accident previously 
evaluated. Since the containment leakage rate will not increase, the 
consequences of any previously evaluated accident will not increase. 
Therefore, the increase in design and peak pressures does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    A structural integrity test (SIT) will be performed at 1.15 
times the new design pressure of 59 psig. The SIT will provide 
acceptance criteria to assure that measured responses are within the 
limits predicted by analyses.
    Additionally, evaluations of components within the containment 
building demonstrate that the components are qualified to the 
increased pressure.
    Revising the allowable containment operating conditions provides 
more operating flexibility than current requirements. The proposed 
change is consistent with the assumptions made in the revised 
containment peak pressure analyses. Since the change only affects 
containment atmosphere conditions allowed during normal operation, 
it has no impact on the probability of initiation of a previously 
evaluated accident. Therefore, this aspect of the change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The increase in peak accident pressure will also require leakage 
rate testing of the containment structure and its penetrations to be 
performed at a 4 psi higher pressure than was required previously. 
Increasing the value of Pa in the containment leakage 
rate program changes the conditions for performing the tests. Since 
the revised value is well within the design capabilities of SSCs 
[systems, structures and components] that could be affected during 
the performance of the test, it will not weaken any of the 
protective barriers. Many past local leak rate tests have been 
performed at increased pressures (59-60 psig) with no significant 
difference in leakage results. Based on the leakage testing history, 
no problems are expected from the increase in Pa. 
Further, since these tests are not performed when the plant is 
operating, they have no impact on normal plant operation or the 
outcome of any previously evaluated accident. Therefore, this aspect 
of the change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Revising the allowable degradation of the containment spray pump 
does not create the probability or consequences of an accident 
previously evaluated. Although the allowable pump degradation 
increased from 6.3% to 10%, analysis has shown that at 10% degraded, 
the pumps can deliver to containment the flow required at 59 psig 
and required to reduce containment pressure to an acceptably low 
level.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From Any Previously Evaluated

    Increasing the containment structural design pressure due [to] 
replacing the steam generators and the future 7-12% power uprate 
does not result in the failure of any system, structure or component 
during the progression of any previously evaluated accident. 
Therefore, the progression of the previously evaluated accidents 
will not change. Further, the change in design pressure is primarily 
administrative and does not affect the way the plant is operated. 
Therefore, this aspect of the change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The added CSAS actuating signal results in isolating the steam 
generators for events that generate a containment pressure high-high 
signal. CSAS, a four channel safety grade system, is part of the 
reactor protection system (RPS). The RPS is designed to reliably 
mitigate the effects of an accident. The only new condition created 
by this change would be the isolation of the steam generators upon 
an inadvertent actuation of CSAS. The possibility of steam generator 
isolation currently exists for an inadvertent MSIS [Main Steam 
Isolation Signal]. This condition is not considered to be an 
accident given the safety grade equipment available to mitigate this 
event and minor consequences due to its occurrence. The CSAS change 
will be implemented such that no new or failure modes or effects 
will be created that could cause a new or different kind of accident 
from any previously evaluated.
    Revising the allowable containment operating conditions permits 
the plant to be operated for a wider range of containment 
atmospheric conditions. This aspect of the proposed change reduces 
the likelihood of a plant upset as a result of shutting the plant 
down in response to exceeding a limiting condition for operation. 
The proposed change is consistent with the assumptions made in the 
accident analysis and will insure that the containment peak pressure 
and temperature do not exceed design limits following design basis 
accidents. Therefore, this aspect of the change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Revising the value of Pa in the containment leakage 
rate program changes the conditions for performing the 10 CFR 50 
Appendix J leak rate test. The revised value is well within the 
design capabilities of SSCs that could be affected during the 
performance of the test. Therefore, this aspect of the change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    Revising the allowable degradation of the containment spray pump 
does not increase the possibility of a new or different kind of 
accident from any previously evaluated. Although the allowable pump 
degradation increased from 6.3% to 10%, analysis has shown that when 
degraded 10%, the pumps can deliver the required flow to the 
containment building at the increased containment pressure of 59 
psig.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    Increasing the containment structural design pressure from 54 to 
59 psig causes a small reduction in the design margin for the 
containment response. Based on the analyses performed, the reduction 
has been

[[Page 9007]]

determined to be acceptable since code allowable stresses are not 
exceeded. The analyses demonstrate that the containment meets all 
applicable codes and standards at 59 psig. Since the physical 
containment structure is not changed as a result of this reanalysis, 
the stresses on the containment structure following a design basis 
event are increased as a result of this change. Since the margin of 
safety is the difference between the stresses that would result in 
containment failure and the stresses at design conditions, this 
change involves a reduction in the margin of safety. However, the 
containment failure pressure is much higher than the design basis 
accident pressure. Also, the DBA [Design Basis Accident] peak 
pressure is currently very close to the design pressure. With the 
proposed change, there is margin between the DBA and design 
pressures. Therefore, this change does not significantly increase 
the probability of containment failure for design basis events. The 
ANO-2 [Arkansas Nuclear One, Unit 2] containment building was 
designed and constructed using significant conservatisms.
    The new application of the CSAS signal is proposed to reduce the 
severity (i.e., reduce the mass and energy addition) of the 
increased effect of a main steam line break inside containment. 
Since this aspect of the proposed change improves the response of 
the plant to this design basis event, it does not involve a 
significant reduction in margin of safety.
    Revising the allowable containment operating conditions provides 
additional operating margin. The proposed allowable operating 
conditions are consistent with the accident analyses performed to 
demonstrate that the peak containment pressure is less than design 
pressure. The relaxation in containment operating conditions was 
made possible by the increase in containment design pressure and the 
addition of the new CSAS actuation to selected components that 
previously received only an MSIS actuation signal.
    Increasing the value of Pa in the containment leakage 
rate program changes the conditions for performing the tests. 
[Fifty-nine] psig is well within the design capabilities [of] SSCs 
that could be affected by the tests. The leakage rate tests will not 
weaken any of the protective barriers. Past local leak rate tests 
have been successfully performed at increased pressures (59-60 psig) 
with no significant difference in leakage results. Therefore, this 
aspect of the change does not involve a significant reduction in the 
margin of safety.
    As discussed previously, increasing the allowable containment 
spray pump degradation does not increase the probability or 
consequences of an accident previously evaluated. Although the 
allowable pump degradation increased from 6.3% to 10%, analysis has 
shown that at 10% degraded, the pumps can deliver the required flow 
to the containment building at the increased containment pressure of 
59 psig.
    Therefore, based on the reasoning presented above and the 
previous discussion of the amendment request, Entergy [Entergy 
Operations, Inc.] has determined that the requested change does not 
involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.
    Attorney for licensee: Nicholas S. Reynolds, Winston and Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 27, 2000.
    Description of amendment request: The proposed amendment would 
delete the current requirements of Technical Specification (TS) 
4.7.9.1.2.d, ``Source installed in the Boronometer,'' associated with 
the installed boronometer sealed source. The source was recently 
removed and stored, and the requirements of TS 4.7.9.1.2.d are no 
longer applicable.
    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:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The modification performed on the boronometer removed its sealed 
source and placed the source in safe storage. The removal of this 
source from plant systems removes the possibility of contamination 
or radiological exposure from this source to personnel working on or 
near the boronometer. Since the source has been placed in safe 
storage, no change in the probability or consequences of an accident 
previously evaluated in evident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From Any Previously Evaluated

    The relocation of the boronometer's sealed source to safe 
storage has not resulted in any new or different kind of accident 
from any previously evaluated. The proposed deletion of 
Specification 4.7.9.1.2.d furthermore does not remove all controls 
from the subject source. While maintained in storage, the 
requirements of Specification 4.7.9.1.2.b will govern testing of the 
sealed source should it be placed in service or transferred to 
another licensee in the future.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    The relocation of the boronometer's sealed source to safe 
storage does not impact the margin to safety. Controls are currently 
established governing sources that are stored and not in use. 
Therefore, deleting the current requirements of Specification 
4.7.9.1.2.d does not result in a reduction in the margin of safety. 
Furthermore, deletion of this surveillance requirement will act to 
reduce radiological exposure to personnel that would normally be 
assigned to perform this activity.

    Therefore, this 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: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.
    Attorney for licensee: Nicholas S. Reynolds, Winston and Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 27, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.4.9.1.2 and delete TS Table 
4.4-5 to remove from the TSs the schedule for the withdrawal of 
reactor vessel material surveillance specimens, pursuant to the 
guidance provided in Generic Letter 91-01, ``Removal of the Schedule 
for the Withdrawal of Reactor Vessel Material Specimens From 
Technical Specifications.'' Changes to the related Bases are also 
proposed.
    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:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the accident conditions and assumptions are not 
affected by the proposed Technical Specification (TS) change. The 
Reactor Vessel Material Surveillance Program ensures the 
availability of data to update the in-service operating temperature 
and pressure limits as well as the Low Temperature Overpressure 
(LTOP) and Pressurized Thermal Shock (PTS) analyses. The schedule 
identifying the

[[Page 9008]]

withdrawal of the surveillance specimens will be removed from the 
TSs; however, the proposed TS 4.4.9.1.2 will continue to require 
that the specimens be removed and examined to determine the changes 
in their material properties, as required by Appendix H to 10CFR50. 
The proposed surveillance specimen removal schedule conforms to ASTM 
[American Society for Testing and Materials] E185-82, ``Standard 
Practice for Conducting Surveillance Tests for Light-Water Cooled 
Nuclear Power Reactor Vessels'' as referenced by 10CFR50, Appendix 
H. No changes to the design of the facility have been made. No new 
equipment has been added or removed and no operational setpoints 
have been altered.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From Any Previously Evaluated

    The proposed change does not add or modify any equipment nor 
does the proposed change involve any operational changes to any 
plant systems or Limiting Conditions for Operation (LCO). As 
required by Appendix H, the proposed change will continue to require 
the specimens be removed and examined to determine changes in their 
material properties. This change does not introduce any new accident 
or malfunction mechanism nor is any physical plant change required.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    Removal of the schedule from Technical Specifications is an 
administrative change and will have no impact on the margin of 
safety. Since changes to the reactor vessel material surveillance 
specimens withdrawal schedule are controlled by the requirements of 
Appendix H to 10CFR50, removing the schedule from Technical 
Specifications will not result in any loss of regulatory control. In 
addition, to ensure the surveillance specimens are withdrawn at a 
proper time, surveillance requirement 4.4.9.1.2 will continue to 
require specimens be removed and examined per the ANO-2 [Arkansas 
Nuclear One, Unit 2] Safety Analysis Report to determine changes in 
their material properties, as required by Appendix H.

    Therefore, this 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: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.
    Attorney for licensee: Nicholas S. Reynolds, Winston and Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 12, 2000 (NPF-38-226).
    Description of amendment request: The proposed change modifies 
Technical Specifications (TS) 3.9.4, ``Containment Building 
Penetrations,'' to allow the containment equipment door, airlocks, and 
other penetrations to remain open, but capable of being closed, during 
core alterations or movement of irradiated fuel in containment. 
Additionally, a note, Bases changes, and Surveillance Requirements 
changes provide further enhancements to clarify equipment door, 
airlock, and penetration closure capability.
    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: The proposed change would allow the containment 
equipment hatch door, personnel air lock (PAL) doors, emergency air 
lock (EAL) doors and penetrations to remain open during fuel 
movement and core alterations. These penetrations are normally 
closed during this time period in order to prevent the escape of 
radioactive material in the event of a fuel handling accident (FHA) 
inside the containment. These penetrations are not initiators of any 
accident. The probability of a FHA is unaffected by the position of 
these penetrations.
    The new FHA analysis with an open containment demonstrates the 
maximum offsite doses are well within the acceptance limits 
specified in SRP [Standard Review Plan] 15.7.4. This FHA analysis 
results in maximum offsite doses of 53.70 rem to the thyroid and 
0.176 rem to the whole body. The calculated control room dose is 
also well within the acceptance criteria specified in GDC [General 
Design Criteria] 19. The analysis results in thyroid and whole body 
dose to the control room operator of 0.932 rem and 0.015 rem, 
respectively.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed change does not involve the addition or 
modification of any plant equipment. Also, the proposed change would 
not alter the design, configuration, or method of operation of the 
plant beyond the standard functional capabilities of the equipment. 
The proposed change involves a change to the Technical 
Specifications (TS) that would allow the equipment hatch door, the 
PAL door, the EAL door and penetrations to be open during core 
alterations and fuel movement within the containment. Having these 
doors and penetrations open does not create the possibility of a new 
accident. Provisions to ensure the capability to close the 
containment will have been made in the event of a FHA.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will the operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response:
    This proposed change has the potential for an increased dose at 
the site boundary due to a FHA; however, the analysis demonstrates 
that the resultant doses are well within the appropriate acceptance 
limits. The margin of safety, as defined by SRP 15.7.4, Rev. 1, has 
not been significantly reduced. The offsite and control room doses 
due to a FHA with an open containment have been evaluated with 
conservative assumptions, such as all airborne activity reaching the 
containment is released instantaneously to the outside atmosphere, 
will ensure the calculation bounds the expected dose. Closing the 
equipment hatch door and at least one door in each personnel airlock 
following an evacuation of the containment reduces the offsite doses 
in the event of a FHA and provides additional margin to the 
calculated offsite doses.
    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.
    Attorney for licensee: N. S. Reynolds, 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: January 25, 2000.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Section 3/4.4.5, ``Reactor Coolant 
System--Steam Generators,'' and its

[[Page 9009]]

associated Bases. In accordance with Framatome Technologies 
Incorporated Topical Report BAW-10236P, Revision 0, ``Addendum for 
Davis-Besse Repair Roll UTS Exclusion Zones,'' the proposed changes 
would modify the repair roll process to update exclusion zones and 
allow the use of the double repair roll for the repair of once-through 
steam generator tubes with defects within the upper tubesheet.
    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, (DBNPS) Unit No. 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because testing and analysis have 
shown the proposed repair roll process to be added to Surveillance 
Requirement (SR) 4.4.5.4.a.7 ensures the new pressure boundary joint 
created by the repair roll process provides structural and leakage 
integrity equivalent to the original design and construction for all 
normal operating and accident conditions. The proposed repair roll 
process does not alter the design or operating characteristics of 
the steam generators or systems interfacing with the steam 
generators. Therefore, the proposed changes to SR 4.4.5.4.a.7 will 
not increase the probability of a previously evaluated accident.
    The proposed change to Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not involve an increase in 
the probability of an accident previously evaluated.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed repair roll 
process to be added to Surveillance Requirement (SR) 4.4.5.4.a.7 
ensures the new pressure boundary joint created by the repair roll 
process provides structural and leakage integrity equivalent to the 
original design and construction for all accident conditions. Should 
a repaired tube fail, the radiological consequences would be bounded 
by the existing Steam Generator Tube Rupture analysis.
    The proposed change to Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not involve an increase to 
the consequences of an accident previously evaluated.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because there will 
be no change in the operation of the steam generators or connecting 
systems as a result of the repair roll process added by the proposed 
changes to SR 4.4.5.4.a.7. The physical changes in the steam 
generators associated with the repair roll process have been 
evaluated and do not create the possibility for a new or different 
kind of accident from any accident previously evaluated, i.e., the 
physical change in the steam generators is limited to the location 
of the primary to secondary boundary within the tubesheet. 
Furthermore, the repair roll process installs a pressure boundary 
joint equivalent to that of the original fabrication. Accordingly, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change to Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not create the possibility 
of any new or different kind of accident.
    3. Not involve a significant reduction in a margin of safety 
because all of the protective boundaries of the steam generator are 
maintained equivalent to the original design and construction with 
tubes repaired by the repair roll process. Furthermore, tubes with 
primary system to secondary system boundary joints created by the 
repair roll have been shown by testing and analysis to satisfy all 
structural, leakage, and heat transfer requirements. The additional 
testing of tubes repaired by the repair roll process under existing 
SR 4.4.5.9 provides continuing inservice monitoring of these tubes 
such that inservice degradation of tubes repaired by the repair roll 
process will be detected. Therefore, the changes to SR 4.4.5.4.a.7 
to modify the repair process do not reduce the margin of safety.
    The proposed change to Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not reduce the margin of 
safety.

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

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

    Date of amendment request: November 30, 1999.
    Description of amendment request: The licensee proposed to amend 
the unit's Technical Specifications (TS), Section 3.4.4, ``Emergency 
Ventilation System [EVS],'' and Section 3.4.5, ``Control Room Air 
Treatment System,'' to require testing consistent with American Society 
for Testing and Materials (ASTM) Standard D3803-1989. The current 
standard specified by these sections is ANSI N510-1980. The licensee's 
application for amendment is a response to the NRC's Generic Letter 
(GL) 99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
    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 operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed TS change will require testing the EVS and Control 
Room Air Treatment System charcoal filters in accordance with ASTM 
D3803-1989 versus ANSI N510-1980. Neither the EVS or Control Room 
Air Treatment System involve initiators or precursors to an accident 
previously evaluated as both systems perform mitigative functions in 
response to an accident. Failure of either system would result in 
the inability to perform its mitigative function but no failure 
would increase the probability of an accident. Accordingly, changing 
the test methodology of the charcoal filters will not affect any 
accident precursors. Therefore, the probability of an accident 
previously evaluated is not increased.
    The NMP1 [Nine Mile Point Unit 1] EVS is designed to limit the 
release of radioactive gases to the environment within the 
guidelines of 1OCFR1OO for analyzed accidents. The Control Room Air 
Treatment System is designed to limit doses to control room 
operators to less than the values allowed by GDC 19. Both systems 
contain charcoal filters which require laboratory carbon sample 
analysis be performed in accordance [with] ANSI [American National 
Standards Institute] N510-1980 as required by TS. Charcoal filter 
samples are tested to determine whether the filter adsorber 
efficiency is greater than that assumed in the design basis accident 
analysis. The proposed TS changes to test the charcoal material in 
accordance with ASTM D3803-1989 (versus ANSI N510) will assure the 
ability of the subject systems to perform their intended function by 
providing a more realistic prediction of the capability of the 
charcoal filters. Therefore, the proposed changes will not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed TS change will require testing the EVS and Control 
Room Air Treatment System charcoal filters in accordance with ASTM 
D3803-1989 versus ANSI N510-1980. This change will not involve 
placing these systems in new configurations or operating the systems 
in a different manner that could result in a new or different kind 
of accident. Testing in accordance with the ASTM D3803-1989

[[Page 9010]]

standard will assure the ability of the subject systems to perform 
their intended function by providing a more realistic prediction of 
the capability of the charcoal filters. Therefore, the proposed 
change will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed TS changes will not adversely affect the 
performance characteristics of the EVS or Control Room Air Treatment 
System nor will it affect the ability of these systems to perform 
their intended functions. Charcoal filter samples are tested to 
determine whether the filter absorber efficiency is greater than 
that assumed in the design basis accident analysis. The proposed TS 
changes to test the charcoal material in accordance with ASTM D3803-
1989 (versus ANSI N510-1980) will assure the ability of the subject 
systems to perform their intended function by providing a more 
realistic prediction of the capability of the charcoal filters. 
Also, the proposed changes are consistent with the changes 
recommended in NRC GL 99-02. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.

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

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment requests: November 10, 1999.
    Description of amendment requests: The proposed amendments would 
modify Technical Specification (TS) 4.12, ``Steam Generator Tube 
Surveillance,'' to revise the elevated F-Star (EF*) distance from 1.62 
inches to 1.67 inches based on Westinghouse Topical Report WCAP-14225, 
Revision 2, entitled ``F* and Elevated F* Tube Plugging Criteria for 
Tube with Degradation in the Tubesheet Region of the Prairie Island 
Units 1 and 2 Steam Generators.'' The change was necessitated by a 
correction of a minor error in the tubesheet bending calculation 
associated with the previously approved EF* criterion. 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[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the EF* Distance ensures the roll 
expansion is sufficient to preclude tube pullout from tube 
degradation located below the EF* Distance, regardless of the extent 
of the tube degradation. The existing Technical Specification 
leakage rate requirements and accident analysis assumptions remain 
unchanged in the unlikely event that significant leakage from this 
region does occur. Tube rupture and pullout is not expected for 
tubes using either the proposed or current EF* Distance because, in 
practice, the roll expanded region exceeds both distances. Any 
leakage out of the tube from within the tubesheet at any elevation 
in the tubesheet is still fully bounded by the existing steam 
generator tube rupture analysis included in the Prairie Island USAR 
[Updated Safety Analysis Report].
    Leakage testing of roll expanded tubes indicates that for roll 
lengths approximately equal to the EF* distance, any postulated 
faulted condition primary to secondary leakage from EF* tubes would 
be insignificant. Leakage testing was previously reported for 2 inch 
effective length hard rolls.
    Thus, neither the probability nor consequences of previously 
evaluated accidents are affected by the proposed increase in the EF* 
Distance.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    Implementation of the proposed EF* Distance does not introduce 
any significant changes to the plant design basis, nor does it 
change the way any system, structure, or component is operated. Use 
of EF* (either using the existing or proposed EF* Distance) does not 
provide a mechanism to initiate an accident outside of the region of 
the expanded portion of the tube. Any hypothetical accident as a 
result of any tube degradation in the expanded portion of the tube 
would be bounded by the existing tube rupture accident analysis.
    Thus, no new or different kind of accident is created by the 
proposed increase in EF* Distance.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The proposed increase in EF* Distance will not decrease the 
integrity of the reactor coolant system boundary. The use of the EF* 
criterion has been previously demonstrated to maintain the integrity 
of the tube bundle commensurate with the requirements of Reg. Guide 
1.121 (intended for indications in the free span of tubes) and the 
primary to secondary pressure boundary under normal and postulated 
accident conditions. Acceptable tube degradation of the EF* 
criterion is any degradation indication in the tubesheet region, 
more than the EF* Distance below the bottom of the transition 
between the roll expansion and the unexpanded tube. The safety 
factors used in the verification of the strength of the degraded 
tube are consistent with the safety factors in the ASME [American 
Society of Mechanical Engineers] Boiler and Pressure Vessel Code 
used in steam generator design.
    The EF* Distance has been verified by testing to be greater than 
the length of roll expansion required to preclude both tube pullout 
and significant leakage during normal and postulated accident 
conditions. Resistance to tube pullout is based upon the primary to 
secondary pressure differential as it acts on the surface area of 
the tube, which includes the tube wall cross-section, in addition to 
the inner diameter based area of the tube. The leak testing 
acceptance criteria are based on the primary to secondary leakage 
limit in the Technical Specifications and the leakage assumptions 
used in the USAR accident analyses.
    Revision of the EF* length does not affect the integrity of the 
existing EF* tubes which are in service due to the conservative 
length of the additional reroll.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the USAR or the Technical Specification 
Bases.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: January 5, 2000.
    Description of amendment request: This request proposes to change 
Technical Specification Section 3/4 6.1.6, including its Bases, and to 
add Section 6.8.4.h. The proposed changes support the new requirements 
of 10 CFR 50.55a, which require licensees to update their Containment 
Vessel Structural Integrity Programs to incorporate the provisions of 
ASME Section XI, Subsection IWL (1992 Edition with 1992 Addenda) and 
the five additional provisions found in 10 CFR 50.55a(b)(2)(viii).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 C.F.R. 50.91(a), the licensee has 
provided its analysis of the

[[Page 9011]]

issue of no significant hazards consideration, which is presented 
below:

    1. This proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes revise the surveillance requirements for 
containment reinforced concrete and unbonded post-tensioning systems 
inservice examinations as required by 10 CFR 50.55a(b)(2)(vi) and 10 
CFR 50.55a(b)(2)(viii). The revised requirements affect the 
inservice inspection program designed to detect structural 
degradation of the containment reinforced concrete and unbonded 
post-tensioning systems and do not affect the function of the 
containment reinforced concrete and unbonded post-tensioning system 
components. The reinforced concrete and unbonded post-tensioning 
systems are passive components whose failure modes could not act as 
accident initiators or precursors.
    The proposed changes do not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility.
    Therefore, this proposed change does not represent a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. This proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new of different 
requirements or introduce a new accident initiator, accident 
precursor or malfunction mechanism. The proposed changes provide an 
NRC approved ASME Code inspection/testing methodology to assure age 
related degradation of the containment structure will not go 
undetected. The function of the containment reinforced concrete and 
unbonded post-tensioning system components are not altered by this 
change. Additionally, there is no change in the types or increases 
in the amounts of any effluent that may be released off-site and 
there is no increase in individual or cumulative occupational 
exposure. Therefore, this proposed change does not create the 
possibility of an accident of a different type than previously 
evaluated.
    3. This proposed change does not involve a significant reduction 
in a margin [of] safety.
    The Reactor Building internal design pressure is 57 psig and the 
maximum peak pressure from a postulated steam line break is 53.5 
psig. The proposed change does not impact the margin of safety 
included in the design pressure compared to the peak calculated 
pressure because the proposed activity does not alter, in any way, 
the available force provided by the tendons. Additionally, the 
proposed activity does not affect the initial temperature conditions 
within the Reactor Building assumed in the accident analysis for a 
steam line break. Therefore, this proposed change does not involve a 
significant reduction in a margin [of] safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: January 27, 2000.
    Description of amendment request: The Virgil C. Summer Nuclear 
Station (VCSNS) Technical Specifications (TS), Section 5.6.1, are being 
revised to replace the maximum reference fuel assembly K infinity 
(K) with a figure of Integral Fuel Burnable Absorbers (IFBA) 
rods per assembly versus nominal fuel enrichment. This change will 
assure that the reactivity requirements for spent fuel storage remain 
satisfied. Additionally, the requirement for new fuel storage is being 
revised to remove K since IFBAs are not considered or 
required in the criticality analysis for new fuel storage.
    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 revise the methodology utilized in 
determining the IFBA requirement for storage of spent fuel. IFBA 
credit is not used in the new fuel storage criticality analysis 
performed by Westinghouse. Removing K infinity (K) from 
these Specifications and replacing the spent fuel requirement with 
the IFBA-enrichment curve will not result in any increase in the 
probability or consequences of an accident previously evaluated. The 
analysis of concern is the criticality analysis for storage of fuel 
in the spent fuel storage racks. The analysis must conclude that 
fuel stored in the configurations allowed in the spent fuel storage 
racks will not result in any unplanned criticality.
    The IFBA rods per assembly versus the nominal enrichment of the 
fuel assembly curve and the K methodology were both 
developed to ensure that Keff in the spent fuel storage 
racks remains less than or equal to 0.95 under all postulated 
conditions. This limit is included in the VCSNS licensing basis. The 
IFBA versus enrichment curve results in determining more accurate 
IFBA requirements than the K methodology, and continues to 
maintain the licensing basis limit.
    This change will not revise the geometry of the spent fuel 
storage racks, the poisons present to prevent criticality, or 
coolant capabilities. The licensing basis limit for reactivity 
control of the spent fuel storage racks remains satisfied.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not result in any change to the design 
or operation of the spent fuel pool or any support systems 
associated with the spent fuel pool. The IFBA requirements developed 
from using the IFBA versus enrichment curve are potentially more 
conservative than developed using the K methodology. There 
are no scenarios that are postulated to occur that would create the 
possibility of a new or different kind of accident from any 
previously evaluated in the FSAR (see original) or FPER (see 
original).
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. IFBA is not assumed in any criticality 
analysis performed for new fuel storage. This change incorporates a 
more accurate method for determining IFBA requirements for fuel 
storage in the spent fuel storage racks. Both the current 
methodology and the proposed methodology have been reviewed and 
approved by the NRC in WCAP-14416-NP-A as acceptable methods for 
assuring that the licensing basis for the spent fuel pool reactivity 
limit remain satisfied. Therefore, the margin of safety with respect 
to unplanned criticality, for the storage of fuel in the spent fuel 
storage racks is not reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

[[Page 9012]]

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

    Date of application for amendments: December 17, 1999 (TS 99-25).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79 
(Unit 2) by modifying License Provision Statement 2.B.(5), in 
conjunction with an exemption to 10 CFR 50.54(ee), to allow temporary 
storage of low-level radioactive waste generated at the Watts Bar 
Nuclear Plant (WBN) at the SQN plant site.
    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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The probability of occurrence or the consequences for an 
accident or malfunction is not increased. Design basis accidents 
were previously analyzed by TVA and reviewed by NRC as part of the 
materials license process for the on-site storage facility (OSF). 
The intended future usage of the OSF is bounded by those analyses, 
with the exception of transport from WBN to SQN. Transport from WBN 
to SQN involves a distance of only 35 miles, which is very likely a 
small increment of the distance to any final off-site repository. 
For example, the 35-mile transit from WBN to SQN is much less than 
the 370-mile distance from WBN to Barnwell, South Carolina. The 
shipment of LLRW from WBN was reviewed as part of the WBN Unit 1 
operating license request (WBN Final Safety Analysis Report [FSAR] 
Section 11.5.6). As with any shipment of low-level radioactive waste 
(LLRW), all Department of Transportation (DOT) requirements will be 
met.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    A possibility for an accident or malfunction of a different type 
than any evaluated previously in SQN's FSAR is not created by the 
proposed change; nor is the possibility for an accident or 
malfunction of a different type. Potential accidents were previously 
analyzed by TVA and reviewed by NRC as part of the materials license 
process for the OSF. The intended future usage of the OSF is bounded 
by those analyses, with the exception of transport from WBN to SQN. 
Radwaste shipments from WBN to SQN will be no different than any 
other radwaste shipment except that the distance is only 35 miles. 
This transportation route does not present any significant potential 
negative impacts on the public health and safety. As with any 
shipment of LLRW, all DOT requirements will be met.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment will not involve a significant reduction 
in the margin of safety. The margin of safety was previously 
analyzed by TVA and reviewed by NRC as part of the materials license 
process for the OSF. The intended future usage of the OSF is bounded 
by those analyses, with the exception of transport from WBN to SQN. 
The transport route from WBN to SQN, which involves a distance of 
only 35 miles, does not present any significant potential negative 
impacts on the public health and safety [and] is very likely a small 
increment of the distance to any final off-site repository. For 
example, this is much less than the distance to Barnwell. The 
shipment of LLRW from WBN was reviewed as part of the WBN Unit 1 
operating license request (WBN FSAR Section 11.5.6). As with any 
shipment of LLRW, all DOT requirements will be met.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: January 13, 2000 (Reference Number TXX-
00011).
    Brief description of amendments: The proposed amendments would 
change the Comanche Peak Steam Electric Station (CPSES) Technical 
Specification (TS) as follows: (1) Revise TS 3.8.3 (Condition B and 
Surveillance Requirement (SR) 3.8.3.2) to conservatively increase the 
required emergency diesel generator (DG) lube oil inventory values, (2) 
revise TS SR 3.8.3.2 to add a note stating that the surveillance is not 
required to be performed until the diesel has been in shutdown greater 
than 10 hours, and (3) delete the footnote associated with SR 3.8.4.7.
    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) Do the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    (a) The proposed changes establish more conservative DG lube oil 
inventory levels to support required DG operations. Conservatively 
revising the required lube oil levels does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (b) The proposed change to add a surveillance note cannot affect 
the probability or consequences of any accident. When surveillances 
are done, it cannot initiate an accident or affect the course of 
mitigation. Lube oil levels are checked after each run. If the lube 
oil level was at the minimum required ``1.75 inches below the low 
static level'' at the start of a normal 24 hour surveillance run, 5 
days of lube oil inventory is provided above the Condition B level 
of ``5.5 inches below the low static level.'' Allowing 10 hours 
after the surveillance run to check the static level is not 
significant because relative lube oil level is maintained during 
engine run through the use of an indicator on the panel ensuring 
adequate oil level during and just after the run. The Condition B 
lube oil inventory ensures a minimum of [2] days of operation before 
any addition of lube oil would be needed. In the event of an 
accident which requires extended run of the emergency diesel 
generators, lube oil can be added with the engines running.
    (c) Deletion of the footnote associated with SR 3.8.4.7, which 
provided a one time exception for the battery surveillance, is an 
administrative change and does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    (a) Plant procedures are only altered to the extent that the 
revised specification will enhance the monitoring of the DG lube oil 
inventory level to support required DG operation at full load 
conditions. These changes ensure continued support of the safety 
related DG, do not involve any physical alteration to the plant, and 
do not affect their failure or failure modes.
    (b) The proposed change to add a surveillance note [does] not 
involve any physical alteration to the plant and [does] not affect 
their failure or failure modes.
    (c) Deletion of the footnote associated with SR 3.8.4.7, which 
provided a one time exception for the battery surveillance, is an 
administrative change and will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Therefore, these changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Do the proposed changes involve a significant reduction in a 
margin of safety?
    (a) The proposed changes will not alter any accident analysis 
assumptions, initial conditions, or results. Conservatively revising 
the required DG lube oil levels will ensure proper DG operations as 
assumed in the safety analyses.
    (b) The proposed change to add a note will not alter any 
accident analysis assumptions,

[[Page 9013]]

initial conditions, or results. Conservatively revising the required 
conditions for DG lube oil level surveillance will ensure proper DG 
operations as assumed in the safety analyses.
    (c) Deletion of the footnote associated with SR 3.8.4.7, which 
provided a one time exception for the battery surveillance, is an 
administrative change and does not involve a significant reduction 
in a margin of safety.
    Therefore, these changes [do] not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: January 13, 2000 (Reference Number TXX-
00010).
    Brief description of amendments: The proposed amendments would 
change Comanche Peak Steam Electric Station (CPSES) Technical 
Specification Surveillance Requirement (SR) 3.3.1.10 to add Note 3 
which would allow entry into Modes 2 or 1 without the performance of N-
16 detector plateau verification until 72 hours after achieving 
equilibrium conditions at greater than or equal to 90% of rated 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:

    (1) Do the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change is considered to be a correction of an 
editorial error. The proposed revision to SR 3.3.1.10 is consistent 
with the current CPSES licensing basis. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed change is considered to be an editorial correction 
and does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    (3) Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed change is considered to be an editorial correction 
and does not involve a reduction in a margin of safety.

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

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

    Date of application request: January 14, 2000 (ULNRC-04172).
    Description of amendment request: The proposed amendment would 
revise several sections of the improved Technical Specification (ITSs) 
to correct eight editorial errors made in either (1) The application 
dated May 15, 1997, (and supplementary letters) for the ITSs or (2) the 
certified copy of the ITSs that was submitted in the licensee's letters 
of May 27 and 28, 1999. The ITSs were issued as Amendment No. 133 by 
the staff in its letter of May 28, 1999, and will be implemented by the 
licensee to replace the current TSs by April 30, 2000. There are no 
changes in any requirements in the ITSs. The proposed changes to the 
ITSs are:
    (1) The correct abbreviation in the table of contents, ITS page 2, 
Section 3.3.7, is ``CREVS'' instead of ``CREFS''.
    (2) The Condition D for limiting condition for operation (LCO) 
3.7.2, ``Main Steam Isolation Valves (MSIVs),'' has a reference to 
itself (Condition D) that should be deleted on ITS page 3.7-5.
    (3) The spelling of ``required'' will be corrected in the 
definition of the Term Actions on ITS page 1.1-1.
    (4) The completion time of 8 hours for Required Action A.2 of 
Example 1.3-6 on ITS page 1.3-10 will be properly relocated to be on 
the same line as A.2.
    (5) The note for Condition D of LCO 3.7.4, ``Atmospheric Steam Dump 
Valves (ASDs),'' on ITS page 3.7-10 will be made the full column width 
of the required action column.
    (6) The word boundary in the note for LCO 3.7.13, ``Emergency 
Exhaust System (EES),'' on ITS page 3.7-31, will not be capitalized.
    (7) The note for Condition A of LCO 3.7.16, ``Fuel Storage Pool 
Boron Concentration,'' on ITS page 3.7-36 will be made the full column 
width of the required action column.
    (8) The colon in 3.1:5 will be replaced by a period to have 3.1.5 
in the list of specifications given in item a.7 of Section 5.6.5, 
``Core Operating Limits Report (COLR),'' on ITS page 5.0-29.
    Basis for proposed no significant 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 changes involve corrections to the ITS that are 
associated with the original conversion application and supplements 
or the certified copy of [the] ITS. The changes are considered as 
administrative changes and do not modify, add, delete, or relocate 
any technical requirements of the Technical Specifications. As such, 
the administrative changes do not effect initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, these changes do 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 changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
changes will not impose any new or eliminate any old requirements.
    Thus, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes will not reduce a margin of safety because 
they have no effect on any safety analyses assumptions. The changes 
are administrative in nature.
    Therefore, the 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

[[Page 9014]]

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

    Date of amendment request: December 21, 1999.
    Description of amendment request: The proposed amendment would 
change Section 15.3 of the Technical Specifications in order to more 
clearly define the requirements for the service water (SW) system 
operability. The December 21, 1999, application supercedes the July 30, 
1998, application that was previously noticed in the Federal Register 
(63 FR 71976) on December 30, 1998.
    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 Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The Service Water System is primarily a support systems required 
to be operable for accident mitigation. Portions of the SW system 
supplying the containment fan coolers also function as part of the 
containment pressure boundary under post accident conditions. 
Failures within the SW system are not an initiating condition for 
any analyzed accident.
    Analyses performed demonstrate that under the Technical 
Specifications allowable configurations, the SW system will continue 
to perform all required functions. The SW system is capable of 
supplying the required cooling water flow to systems required for 
accident mitigation. That is, the SW system removes the required 
heat from the containment fan coolers and residual heat removal heat 
exchangers ensuring containment pressure and temperature profiles 
following an accident are as evaluated in the FSAR [Final Safety 
Analysis Report]. This in turn ensures that environmental 
qualification of equipment inside containment is maintained and thus 
functions as required post-accident.
    SW system response post accident is within all design limits for 
the system. Transient and steady state forces within the system 
remain within all design and operability limits, thereby maintaining 
the integrity of the system inside containment and the integrity of 
the containment pressure boundary. Assumptions dependent on the 
containment pressure profile for containment leakage assumed in the 
radiological consequences analyses remain valid.
    In addition, removing required heat from containment ensures 
that cooling of the reactor core is accomplished for long-term 
accident mitigation.
    Therefore, operation of the SW system as proposed will not 
result in a significant increase in the probability or consequences 
of any accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes do not alter the way in which the SW system 
performs its design functions nor the design criteria of the system. 
The proposed changes do not introduce any new or different normal 
operation or accident mitigation functions for the system. 
Therefore, no new accident initiators are introduced by the proposed 
changes. Operation of [the] SW system as proposed cannot result in a 
new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    Analyses performed in support of the proposed amendments 
demonstrate that the SW system continues to perform its function as 
assumed and credited in the accident analyses and radiological 
consequence analyses performed for the Point Beach Nuclear Plant. 
The SW flow analyses conservatively assume limiting calculational 
parameters such as minimum allowed IST [inservice testing] pump 
performance curves, minimum credible pump bay level, maximum 
postulated lake temperature, inclusion of system water leakage, 
maximum flow through system temperature control valves, bounding 
values for system throttle valve settings and impacts of instrument 
inaccuracy. Therefore, the analyses and results are not changed. All 
analysis limits for the system remain met. The SW system continues 
to be operated and responds within all design limits for the system. 
Therefore, operation of the Point Beach Nuclear Plant in accordance 
with the proposed amendments cannot result in a significant 
reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of application for amendment: December 28, 1999.
    Brief description of amendment: The amendment would revise the 
reactor vessel material coupon withdrawal schedule specified in 
Technical Specifications Table 4.4.6.1.3-1, entitled ``Reactor Vessel 
Material Surveillance Program-Withdrawal Schedule.''
    Date of publication of individual notice in Federal Register: 
January 14, 2000 (65 FR 2443).
    Expiration date of individual notice: February 14, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances

[[Page 9015]]

provision in 10 CFR 51.12(b) and has made a determination based on that 
assessment, it is so indicated.
    For further details with respect to the action see: (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

Arizona Public Service Company, et al., Docket 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: February 26, 1999, as 
supplemented May 21, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications to extend the completion time for one 
inoperable low pressure safety injection subsystem from 72 hours to 7 
days. These amendments provide partial response to the licensee's 
application for amendments. The remaining request will be addressed 
under separate correspondence.
    Date of issuance: February 1, 2000.
    Effective date: February 1, 2000, to be implemented within 45 days.
    Amendment Nos.: Unit 1--124, Unit 2--124, Unit 3--124.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17023).
    The May 21, 1999, supplement provided clarifying information that 
was within the scope of the original Federal Register notice and did 
not change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 1, 2000.
    No significant hazards consideration comments received: No.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: August 27, 1999, as 
supplemented September 20, 1999.
    Brief description of amendments: The amendments would modify the 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Technical 
Specifications to allow placement of one or more assemblies on spent 
fuel rack spacers to support fuel reconstitution activities in the 
spent fuel pool.
    Date of issuance: February 3, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 233 and 209.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 22, 1999 (64 
FR 51345).
    The September 20, 1999, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated February 3, 2000.
    No significant hazards consideration comments received: No.

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 application for amendment: August 4, 1999, as supplemented 
December 3, 1999, and January 11, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification 6.9.1.6.2 to incorporate analytical methodology 
references which are used to determine core operating limits. The 
analytical methodologies referenced are documented in topical reports 
which have been accepted by the Nuclear Regulatory Commission for 
referencing in licensing applications.
    Date of issuance: February 10, 2000.
    Effective date: February 10, 2000.
    Amendment No.: 94.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46426).
    The December 3, 1999, and January 11, 2000, submittals contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 2000.
    No significant hazards consideration comments received: No.

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: November 12, 1999, as 
supplemented by letter dated January 10, 2000.
    Brief description of amendments: The amendments changed Technical 
Specification (TS) 3/4.6.K to revise the reactor pressure boundary 
pressure-temperature limits, changed TS 3/4.12.C to delete a special 
test exception which allows performance of the hydrostatic test above 
212 degrees Fahrenheit while in Mode 4, and changed TS 3/4.6.P to 
clarify the operability requirements for the residual heat removal 
system during the hydrostatic test.
    Date of issuance: February 4, 2000.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 195 & 191.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70081).
    The January 10, 2000, letter 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 February 4, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 29, 1999 as supplemented by 
letter dated October 20, 1999.
    Brief description of amendment: The amendment consists of changes 
to Surveillance Requirements (SR) 3.8.4.6 of Technical Specifications 
(TS) 3.8.4, ``DC Sources--Operating'' and SR 3.8.5.1 of TS 3.85, ``DC 
Sources--Shutdown.''
    Date of issuance: January 28, 2000.
    Effective date: January 28, 2000, and shall be implemented within 
30 days from the date of issuance.
    Amendment No.: 160.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46432).
    The October 20, 1999, supplemental letter corrected the page 
numbering of the technical specifications and did not expand the scope 
of the application as originally noticed and did not change the staff's 
original no significant hazards consideration determination.

[[Page 9016]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 25, 1999, as supplemented by 
letter dated December 9, 1999.
    Brief description of amendment: The amendment consists of a 
modification to TS 3/4.5.1 to allow up to 72 hours to restore safety 
injection tank (SIT) operability if one SIT is inoperable due to boron 
concentration not within the limits or the inability to verify level or 
pressure. The proposed change also allows up to 24 hours to restore SIT 
operability if one SIT is inoperable due to other reasons when reactor 
coolant system pressure is greater than or equal to 1750 pounds per 
square inch, absolute.
    Date of issuance: February 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 155.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9191).
    The December 9, 1999, letter provided additional information that 
did not change the scope of the application as initially noticed or 
change proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 16, 1998, as 
supplemented by letter dated May 10, 1999, and December 8, 1999.
    Brief description of amendment: This amendment changes portions of 
the Technical Specifications regarding the Service Water System.
    Date of issuance: February 3, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 89.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66596).
    The May 10 and December 8, 1999, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 3, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 16, 1999.
    Brief description of amendment: The amendment relocates Technical 
Specification (TSs) 3/4.9.3.2, ``Refueling Operations, Spent Fuel 
Temperature,'' 3/4.9.3.3, ``Refueling Operations, Decay Time,'' 3/
4.9.5, ``Refueling Operations, Communications,'' 3/4.9.6, ``Refueling 
Operations, Crane Operability--Containment Building,'' and 3/4.9.7, 
``Refueling Operations, Crane Travel--Spent Fuel Storage Building,'' to 
the Millstone, Unit No. 2 Technical Requirements Manual. The associated 
Bases pages and index pages are also modified to address the proposed 
change.
    Date of issuance: February 10, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 240.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54378).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 2000.
    No significant hazards consideration comments received: No.

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: February 19, 1998, as 
supplemented July 28, 1999.
    Brief description of amendment: The amendment implements the 
Radioactive Effluent Technical Specifications and makes changes 
necessary to implement the revised 10 CFR Part 20.
    Date of issuance: February 7, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 199.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46442).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 2000.
    No significant hazards consideration comments received: No.

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: October 16, 1998, as 
supplemented January 28, 1999.
    Brief description of amendment: The amendment relocates the 
Chemical and Volume Control System Technical Specifications.
    Date of issuance: February 7, 2000.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 200.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9200).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 6, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification Definition 1.9, ``Core Alterations,'' to explicitly 
define core alterations as the movement of any fuel, sources, or 
reactivity control components within the reactor vessel with the vessel 
head removed and fuel in the vessel.
    Date of issuance: February 1, 2000.
    Effective date: February 1, 2000, to be implemented within 30 days.
    Amendment Nos.: Unit 1-123; Unit 2-111.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.

[[Page 9017]]

    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73099).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 1, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 12, 1999.
    Brief description of amendments: These amendments revise Technical 
Specification Section 3.9.4.c, ``Containment Building Penetrations,'' 
and the associated bases to allow use of administrative controls to 
unisolate certain containment penetrations during refueling operations.
    Date of issuance: February 11, 2000.
    Effective date: As of date of issuance to be implemented no later 
than 45 days after issuance.
    Amendment Nos.: 249 and 240.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1928).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 11, 2000.
    No significant hazards consideration comments received: No.

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: June 22, 1999, as supplemented 
December 17, 1999.
    Brief description of amendments: The amendments reflect changes to 
the Technical Specifications in order to incorporate the Westinghouse 
422V+ fuel assemblies into the reactor cores.
    Date of issuance: February 8, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 193 and 198.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40910).
    The December 17, 1999, letter provided clarifying information that 
was 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 February 8, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 16th day of February, 2000.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-4236 Filed 2-22-00; 8:45 am]
BILLING CODE 7590-01-P