[Federal Register Volume 65, Number 17 (Wednesday, January 26, 2000)]
[Notices]
[Pages 4268-4295]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-1732]



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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 December 31, 1999, through January 14, 2000. 
The last biweekly notice was published on January 12, 2000.

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
of 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 February 25, 2000, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such

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a supplement which satisfies these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: December 17, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 2.1.1.2 to incorporate cycle-specific 
safety limit minimum critical power ratios (SLMCPRs) for the core that 
will be loaded during the upcoming refueling outage
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed license amendment establishes a revised SLMCPR 
value of 1.07 for two recirculation loop operation and 1.09 for 
single recirculation loop operation. The derivation of the cycle-
specific SLMCPRs was performed using NRC approved methods and 
uncertainties described in Amendment Number 25 to NEDE-24011-P-A 
(GESTAR II) and Licensing Topical Reports NEDC-32601P-A, 
``Methodology and Uncertainties for Safety Limit MCPR Evaluations'' 
and NEDC-32694P-A, ``Power Distribution Uncertainties for Safety 
Limit MCPR Evaluation.''
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established, consistent with NRC 
approved methods, to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable.
    The probability of an evaluated accident is not increased by 
revising the SLMCPR values. The change does not require any physical 
plant modifications or physically affect any plant components. 
Therefore, no individual precursors of an accident are affected.
    The proposed license amendment establishes a revised SLMCPR that 
ensures that the fuel is protected during normal operation and 
during any plant transients or anticipated operational occurrences. 
Specifically, the reload analysis demonstrates that a SLMCPR value 
of 1.07 (1.09 for single loop operation) ensures that less than 0.1 
percent of the fuel rods will experience boiling transition during 
any plant operation if the limit is not violated.
    Based on (1) the determination of the new SLMCPR values using 
NRC approved methods and uncertainties, and (2) the operability of 
plant systems designed to mitigate the consequences of accidents not 
having been changed; the consequences of an accident previously 
evaluated have not been increased.
    Therefore, the proposed Technical Specification change does not 
involve an 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 license amendment involves a revision of the SLMCPR 
from 1.11 to 1.07 for two recirculation loop operation and from 1.13 
to 1.09 for single loop operation based on the results of analysis 
of the Cycle 8 core which will once again be fully loaded with GE11 
fuel. Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in the 
allowable methods of operating the facility. This proposed license 
amendment does not involve any modifications of the plant 
configuration or changes in the allowable methods of operation. 
Therefore, the proposed Technical Specification change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed license amendment establishes a revised SLMCPR 
value of 1.07 for two recirculation loop operation and 1.09 for 
single recirculation loop operation. The derivation of these revised 
SLMCPRs was performed using NRC approved methods and uncertainties 
described in Amendment Number 25 to NEDE-24011-P-A (GESTAR II) and 
Licensing Topical Reports NEDC-32601P-A, ``Methodology and 
Uncertainties for Safety Limit MCPR Evaluations'' and NEDC-32694P-A, 
``Power Distribution Uncertainties for Safety Limit MCPR 
Evaluation.'' Use of these methods ensures that the resulting SLMCPR 
satisfies the fuel design safety criteria that less than 0.1 percent 
of the fuel rods experience boiling transition if the safety limit 
is not violated. Based on the assurance that the fuel design safety 
criteria will be met, the proposed license amendment does not 
involve a significant reduction in a margin of safety.

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

[[Page 4270]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Section Chief: Claudia M. Craig.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: December 17, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement (SR) 3.6.1.3.9 
to relax the SR frequency by allowing a representative sample of excess 
flow check valves (EFCVs) to be tested every 18 months, such that each 
EFCV will be tested at least once every 10 years. Current SR 3.6.1.3.9 
requires all EFCVs to be tested every 18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The current SR frequency requires each reactor instrumentation 
line EFCV to be tested every 18 months. The EFCVs at Fermi 2 are 
designed to close automatically in the event of a line break 
downstream of the valve. Indicating lights on a control room panel 
monitor EFCV positions. These valves may be reopened by actuation of 
a solenoid valve, which is operated from a local control panel. 
EFCVs at Fermi 2 are designed and installed following the guidance 
of Regulatory Guide 1.11. This proposed change allows a reduced 
number of EFCVs to be tested every 18 months. Industry operating 
experience, documented in BWROG [Boiling Water Reactor Owners Group] 
Report B21-00658-01, concludes that a change in surveillance test 
frequency has a minimal impact on the reliability for these valves. 
A failure of an EFCV to isolate cannot initiate previously evaluated 
accidents; therefore, there can be no increase in the probability of 
occurrence of an accident as a result of this proposed change.
    Fermi 2 UFSAR [Updated Final Safety Analysis Report], Subsection 
15.6.2 evaluates an instrument line pipe break within secondary 
containment. The evaluation assumes that a small instrument line 
instantaneously and circumferentially breaks at a location where it 
may not be possible to isolate it and where immediate detection is 
not automatic or apparent. The evaluation concluded that 
pressurization of the secondary containment would not result from an 
instrument line break and a failure of the associated EFCV to 
isolate the ruptured line. The standby gas treatment system is not 
impaired by this event, and the calculated offsite exposure is 
substantially below the guidelines of 10 CFR 100. Additionally, 
coolant lost from such a break is inconsequential when compared to 
the makeup capabilities of the feedwater or RCIC [reactor core 
isolation cooling] system. The BWROG report concludes that the risk 
to the public with the extended testing interval is several orders 
of magnitudes below the general public annual exposure limits in 10 
CFR 20.105.
    Although not expected to occur as a result of this change, the 
postulated failure of an EFCV to isolate as a result of reduced 
testing is bounded by the analysis in the UFSAR. Therefore, there is 
no increase in the previously evaluated consequences of the rupture 
of an instrument line and there is no potential increase in the 
radiological consequences of an accident previously evaluated as a 
result of this change.
    2. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change allows a reduced number of EFCVs to be 
tested each operating cycle. No other changes in requirements are 
being proposed. Industry operating experience as documented in the 
BWROG report provides supporting evidence that the reduced testing 
frequency will not affect the high reliability of these valves. The 
potential failure of an EFCV to isolate as a result of the proposed 
reduction in test frequency is bounded by the evaluation of an 
instrument line pipe break described in Subsection 15.6.2 of the 
UFSAR. This change is not a physical alteration of the plant and 
will not alter the operation of the structures, systems and 
components as described in the UFSAR. Therefore, a new or different 
kind of accident will not be created.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The consequences of a postulated instrument line pipe break have 
been evaluated in Subsection 15.6.2 of the UFSAR. The evaluation 
assumed the line instantaneously and circumferentially breaks at a 
location where it may not be possible to isolate it and that the 
EFCV fails to isolate the break. Therefore, any potential failure of 
an EFCV as a result of the reduced testing frequency is bounded by 
this evaluation and 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: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Section Chief: Claudia M. Craig

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: August 18, 1999.
    Description of amendment request: The proposed change would 
amend Technical Specification 3.5.3 and its associated Bases to 
reflect a change in the reactor coolant system (RCS) low pressure 
setpoint for Arkansas Nuclear One, Unit 1 (ANO-1). The RCS low 
pressure setpoint has been adjusted in the conservative direction to 
account for both the uncertainties associated with the actual value 
and the current number of plugged steam generator tubes.
    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:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion of 
these standards as they relate to this amendment request follows:
Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated
    The proposed change to raise the current technical specification 
(TS) ESAS [Engineered Safeguards Actuation Signal] setpoint for low RCS 
pressure does not require new hardware or physical equipment 
modifications to the plant design. By raising the setpoint, a more 
prompt actuation of associated safeguards equipment will be achieved 
for the accident scenarios previously analyzed in the ANO-1 Safety 
Analyses Report (SAR). A more expeditious actuation will ensure a more 
timely response to the accident and serve to potentially decrease the 
consequences of an accident. The RCS Pressure LO LO [Low Low] alarm 
setpoint has also been raised and applicable procedures revised to 
provide the operator sufficient time to bypass the actuation during 
controlled plant maneuvers.
    Therefore, the raising of the low RCS pressure ESAS setpoint from 
1526 psig [pounds per square inch, guage] to 1585 psig 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 is relevant to accident response and mitigation 
and has no [a]ffect on accident initiation. An inadvertent actuation of 
the HPI [high pressure injection] system could result in pressurizing 
the RCS to the point where a pressurizer safety valve could open and 
subsequently fail to close, resulting in a loss of coolant accident.

[[Page 4271]]

However, this event remains unaffected for normal power operations and 
requires discussion of depressurization events only, such as a planned 
cooldown, when an inadvertent actuation could occur earlier due to the 
proposed higher setpoint. This concern is mitigated by the increase of 
the RCS Pressure LO LO alarm setpoint from approximately 1550 psig to 
1640 psig, thus providing the operator ample time to bypass the low RCS 
pressure ESAS setpoint prior to inadvertent actuation. Therefore, no 
new, previously unevaluated event has been introduced relating to the 
inadvertent actuation of HPI components due to the proposed change.
    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 proposed change conservatively raises the existing low RCS 
pressure ESAS setpoint to a new value using existing installed 
equipment. The new value provides protection for the entire spectrum of 
break sizes based on applicable evaluations and considers the effects 
of projected steam generator tube plugging activities. The setpoint is 
also sufficiently below normal operating pressure to aid in preventing 
spurious initiation.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Therefore, based on the reasoning presented above and the previous 
discussion of the amendment request, Entergy Operations, Inc. has 
determined that the requested change does not involve a significant 
hazards consideration.

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

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: November 16, 1999.
    Description of amendment request: The proposed changes to the 
Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Technical 
Specifications (TSs) and associated Bases would provide a 30-day 
allowable outage time (AOT) for Startup Transformer No. 2 (SU#2) which 
is an offsite power source shared by both units. This 30-day AOT would 
be used infrequently for the purpose of performing preventative 
maintenance on the transformer to increase its reliability. The current 
TS constraints would require both units to be in cold shutdown in order 
to perform this maintenance. In addition, changes have been requested 
to the requirements associated with demonstrating the operability of 
the emergency diesel generators to increase the reliability of this 
power supply.
    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:

An evaluation of the proposed changes has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:

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

    Based on existing methodologies, guidance, and procedures 
utilized at ANO, including required assessments of risk associated 
with any significant maintenance activity, the provision of a 30-day 
AOT for preplanned preventative maintenance on SU#2 is acceptable. 
The resulting increase in overall risk was considered to fall into 
NRC Risk Region III (``Very Small Change''). Additionally, removal 
of SU#2 from service in any plant mode of operation has been 
previously evaluated and found acceptable given the existing 
guidance and regulations associated with offsite power sources.
    Five offsite power feeds are available to the ANO switchyard 
with no more than two of the feeds in close proximity to one another 
for a given length, except within the switchyard itself. Failure of 
one feed, regardless of the cause, will result in no more than one 
additional failure, leaving at least three offsite power sources yet 
available, assuming the failure remains outside the ANO switchyard. 
For events that pose a threat within the ANO switchyard, four 
redundant Class 1E EDGs [emergency diesel generators] and one 
Alternate AC [alternating current] diesel generator are capable of 
supply power to the units. Upon loss of the remaining offsite power 
transformer of a unit which may be off-line, offsite power may be 
restored via backfeed operations from the Main Transformers to the 
Unit Auxiliary transformer to supply in-house loads. This ensures 
the availability of redundant power sources, including the 
applicable contingencies established during safety-related equipment 
maintenance performed at ANO, are sufficient in maintaining safe 
unit operations during preplanned preventative maintenance on SU#2 
transformer. Therefore, providing a 30-day AOT for preplanned 
preventative maintenance on SU#2, not to be applied more than once 
in any 10-year period, does not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The elimination of excessive EDG operability demonstrations 
(cold starts) during periods when another required power source is 
inoperable acts to enhance overall EDG reliability and is consistent 
with guidance provided in NRC Generic Letter 84-15 ``Proposed Staff 
Actions to Improve and Maintain Diesel Generator Reliability'' and 
the Revised Standard Technical Specifications (NUREG-1430 and -
1432). Verification of the operability of the remaining EDG will be 
performed within 24 hours should the failure mechanism that caused 
the inoperability of the redundant EDG be concluded to be a common 
cause type failure. The start test in the latter case acts to ensure 
that an EDG source remains available when the cause of the failure 
of the redundant EDG might impact the remaining EDG.
    Therefore, eliminating excessive EDG cold starts 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 removal of SU#2 from service to support needed maintenance 
activities has been previously evaluated for all modes of plant 
operation. Extending the current AOT to 30 days on a limited basis 
does not result in any new accident initiator. The EDGs are not 
considered accident initiators, but are designed to support 
mitigation of accident scenarios. The elimination of excessive EDG 
cold starts acts to enhance overall EDG reliability and has no 
effect on accident development.
    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 associated probabilistic risk assessments indicate that the 
proposed 30-day AOT for SU#2 does not involve a significant increase 
in overall site risk, nor reduce the margin to safety. Thorough 
contingency action planning, which acts to maintain the operability 
of other equipment important to safety during the SU#2 maintenance 
window, additionally acts to ensure the margin to safety is 
maintained. The EDGs are important to safety in that they are 
designed to supply power to safety system components and equipment 
during a loss of offsite power. The elimination of excessive cold 
starts of the EDGs acts to enhance the overall reliability of the 
EDGs and, therefore, proper mitigation of accident scenarios is 
likewise enhanced.

[[Page 4272]]

    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based on the reasoning presented above and the 
previous discussion of the amendment request, 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.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston 
and Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 16, 1999.
    Description of amendment request: The proposed license amendment 
request would revise Fuel Handling Accident (FHA) dose calculations 
for 3 scenarios documented in the River Bend Station, Unit 1 (RBS), 
Updated Safety Analysis Report (USAR). The first is a FHA in the 
fuel building, assumed to occur 24 hours post-shutdown. A second FHA 
analysis was prepared to support Amendment 35 to RBS Technical 
Specifications (TS) which assumed a FHA occurs in the primary 
containment 80 hours post-shutdown during Local Leakage Rate Testing 
(LLRT). A third analysis was prepared in support of Amendment 85 to 
the RBS TS which assumed the containment is open at 11 days.
    These analyses are being updated to account for several changes. 
The primary reason for the revisions, as stated by the licensee, was 
to update the analyses to reflect current RBS operating strategies 
and make the analyses consistent with each other. Specifically, 
Cases 1 and 2 of the three analyses assumed a Radial Peaking Factor 
(RPF) of 1.5 consistent with Regulatory Guide (RG) 1.25. However, 
current core design strategies could lead to an RPF as high as 1.65. 
In addition, to account for the potential impact of extended burnup 
fuel in future operating cycles, an increased iodine-131 gap 
fraction of 0.12 was more conservatively assumed in lieu of the 0.10 
recommended by RG 1.25. The revised analysis also includes a change 
to the control room atmospheric dispersion factors (/Q) 
for the Main Control Room (MCR) ventilation system. Credit is taken 
for Standard Review Plan (SRP) Section 6.4 guidance for manual dual 
control room air intakes in that the /Q's are divided by 
4. The revised FHA analyses also credit this action at a 20 minute 
delay to be consistent with the Loss of Coolant Accident (LOCA) 
analysis.
    Furthermore, an error was discovered in one of the FHA 
calculations. The release rate assumed in the analysis did not 
ensure that the RG 1.25 assumption of a 2-hour release was 
preserved. The error is the result of an inherent bias in the 
secondary mixing effects in the dose calculation. The results 
continue to be bounded by the guidance contained in SRP 15.7.4 and 
RG 1.25.
    Reanalysis showed that the release rate error, compounded with 
the other changes discussed above, resulted in calculated doses 
greater than those currently found in the RBS USAR. In addition, 
some of the doses were also greater than those presented in the 
Amendment 85 submittal. However, the licensee has stated that the 
results of the revised analyses remain ``well within'' 10 CFR 100, 
the guidance contained in SRP 15.7.4, and RG 1.25. Since the 
analyses results are above those reported in the RBS USAR, the 
criterion of 10 CFR 50.59(a)(2)(i) is, therefore, satisfied. 
Accordingly, the licensee has concluded that these changes involve 
an unreviewed safety question.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The analyses changes described by this proposed change to the 
USAR are not initiators to events, and, therefore, do not involve 
the probability of an accident. The changes to the FHA calculations 
for radiological doses following a FHA reflect the current operating 
strategies and make the analyses consistent. These changes included:
     accounting for the impact of extended burnup fuel,
     addressing a change to the control room atmospheric 
dispersion factors assumed in the analysis, and
     revising the Radial Peaking Factor (RPF) used in the 
analysis. Current core design strategies could lead to a RPF higher 
[than] that assumed in Regulatory Guide 1.25.
    The TRANSACT code is used for offsite dose and control room dose 
calculations. The TRANSACT code is derived from the TACT V code 
documented in NUREG/CR-5109. RBS has benchmarked the TRANSACT code 
as discussed in the request dated August 17, 1995, (RBG-41728) which 
resulted in the NRC granting Amendment 85.
    The revisions to the FHA are used to establish operational 
conditions where specific activities represent situations where 
significant radioactive releases can be postulated. These 
operational conditions include:
     initial fuel movement in the Fuel Building 24 hours 
after shutdown,
     fuel movement in Primary Containment after 80 hours 
with leakrate testing being conducted, and
     fuel movement in Primary Containment with the Primary 
Containment open.
    Because the analyses affected by the changes are not considered 
an initiator to any previously analyzed accident, these changes 
cannot increase the probability of any previously evaluated 
accident. Therefore, this change does not increase the probability 
of occurrence of an accident evaluated previously in the safety 
analysis report (SAR).
    This proposed change to the USAR does increase the consequences 
of an accident, but the increase is within all regulatory limits and 
guidance. While the calculated off-site and control room doses of a 
FHA did increase, the dose consequences remain below the regulatory 
limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC [General Design 
Criterion]-19 as approved per NUREG-0989, and the guidance contained 
in SRP 15.7.4 of less than 25% of the 10 CFR 100 limits. The cause 
of these events remains the failure of the fuel assembly lifting 
mechanism. These analyses demonstrated that for the worst case 
bundle drop, the regulatory dose guidelines of SRP 15.7.4 continue 
to be satisfied for the required decay periods.
    This change accounts for the potential effects of current fuel 
design and operating strategies including increased burnup of fuel, 
increased iodine-131 fraction released, Main Control Room 
ventilation system operation, and release rate timing assumptions. 
Reanalysis of the off-site dose calculation demonstrates that the 
revised doses are increased but remain less than the regulatory 
limits of 10 CFR 100 and within the guidance of SRP 15.7.4. 
Therefore, this change does not significantly increase the 
consequences of an accident previously evaluated in the SAR.
    The proposed changes, in conjunction with existing 
administrative controls, bound the conditions of the current design 
basis fuel handling accident analysis. The analysis also concludes 
the limiting offsite radiological consequences are well within the 
acceptance criteria of NUREG[-]0800, Section 15.7.4 and 10 CFR 50, 
Appendix A, GDC[-]19. The analysis is also conducted in a 
conservative manner containing margins in the calculation of 
mechanical analysis, iodine inventory, and iodine decontamination 
factor. Each of these conservatisms will further decrease the 
consequences. Therefore, the proposed changes do not significantly 
increase the probability or consequences of any previously evaluated 
accident.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous[ly] analyzed.

[[Page 4273]]

    This change does not involve initiators to any events in the 
SAR, nor does the activity create the possibility for any new 
accidents. Rather, this change is a result of the evaluation of the 
most limiting FHA, which can occur at River Bend.
    The proposed changes to the dose analyses are consistent with 
previous limits, only revising previous evaluations to account for 
current operating strategies and assumptions. These changes 
included:
     accounting for the impact of extended burnup fuel,
     addressing a change to the control room atmospheric 
dispersion factors assumed in the analysis, and
     revising the Radial Peaking Factor (RPF) used in the 
analysis. Current core design strategies could lead to a RPF higher 
[than] that assumed in Regulatory Guide 1.25.
    The radiological consequences remain within accepted limits of 
10 CFR 100 and guidance of the Standard Review Plan (NUREG-0800) 
Section 15.7.4. Therefore, these changes are consistent with the 
design basis analysis. The proposed changes do not introduce any new 
modes of plant operation and do not involve physical modifications 
to the plant. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previous[ly] analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The dose consequences are calculated in accordance with 
regulatory guidance found in Regulatory Guide 1.25 and the SRP 
[S]ection 15.7.4. The RBS analyses conservatively assumed that 
failures are consistent with those in the standard General Electric 
GESTAR II. These analyses result in a bounding number of fuel 
failures. The RBS analyses are also consistent with those approved 
by the NRC [Nuclear Regulatory Commission] in support of Technical 
Specification Amendments 35 and 85 to the River Bend Station license 
(NPF-47). The radiological dose consequences resulting from these 
failures are therefore analyzed using accepted methods and criteria. 
In addition, the analyses contain known conservatisms and margins to 
ensure the results will remain bounding.
    The revised limits are used to establish operational conditions 
where specific activities represent situations where significant 
radioactive releases can be postulated. These operational conditions 
are consistent with the design basis analysis and are established 
such that the radiological consequences are at or below the current 
regulatory limits and guidance. Safety margins and analytical 
conservatisms have been evaluated and are well understood. 
Conservative methods of analysis are maintained through the use of 
accepted methodology and benchmarking the proposed methods to 
previous analysis. Margins are retained to ensure that the analysis 
adequately bounds all postulated event scenarios. The proposed 
change only eliminates some excess conservatism from the analysis.
    In addition, EOI [Entergy Operations, Inc.] has implemented 
NUMARC [Nuclear Management and Resources Council (now NEI)] 91-06 
guidelines for shutdown operations at RBS. Shutdown Operations 
Protection Plan and Primary-Secondary Containment Integrity 
procedures presently include guidance for closure of the containment 
hatch and other significant openings in containment, in addition to 
the requirements contained in the license and design basis. This 
additional protection will enhance the ability to limit offsite 
effects.
    Acceptance limits for the fuel handling accident are provided in 
10 CFR 100 with additional guidance provided in NUREG[-]0800, 
Section 15.7.4. The proposed changes continue to ensure that the 
whole-body and thyroid doses at the exclusion area and low 
population zone boundaries, as well as control room doses, are below 
the corresponding regulatory limits. These margins are unchanged, 
therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The commission has provided guidance concerning the application 
of the standards of 10 CFR 50.92 by providing certain examples (51 
FR 7751, March 6, 1986) of amendments that are not considered likely 
to involve a significant hazards consideration. This proposed 
amendment is very similar to example (vi):
    (vi) A change which either may result in some increase to the 
probability or consequences of a previously-analyzed accident or may 
reduce in some way a safety margin, but where the results of the 
change are clearly within all acceptable criteria with respect to 
the system or component specified in the Standard Review Plan: for 
example, a change resulting from the application of a small 
refinement of a previously used calculational model or design 
method.
    As we have shown in the preceding discussion, this refinement to 
the FHA dose calculation results in a small increase to the 
consequences of a previously analyzed accident, but the results of 
the change remain clearly within the guidelines of 10 CFR 100, 
Appendix A, GDC[-]19, and the guidance of SRP [S]ection 15.7.4, 
without reducing 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 Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 20, 1999.
    Description of amendment request: The proposed amendment would 
change River Bend Station (RBS) Technical Specification (TS) 3.6.1.3, 
``Primary Containment Isolation Valves (PCIVs),'' to allow the Inclined 
Fuel Transfer System (IFTS) primary containment isolation blind flange 
to be removed during MODE 1, 2, or 3. In its application, the RBS 
licensee stated that, with the blind flange removed and certain 
restrictions and administrative controls in place, the IFTS penetration 
would not represent an uncontrolled breach of the containment boundary 
and that the containment isolation function would continue to be 
provided through implementation of these additional controls.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change permits removal of the blind flange on the 
Inclined Fuel Transfer System (IFTS) when primary containment 
operability is required in MODE[S] 1, 2, and 3. This will permit 
operation of the IFTS while the plant is operating. With respect to 
the probability of an accident, this aspect of the containment 
structure does not directly interface with the reactor coolant 
pressure boundary. The removal of this blind flange does not involve 
modifications to plant systems or design parameters that could 
contribute to the initiation of any accidents previously evaluated. 
Operation of IFTS is unrelated to the operation of the reactor, and 
there is no aspect of IFTS operation that could lead to or 
contribute to the probability of occurrence of an accident 
previously evaluated. Removal of the blind flange and operation of 
IFTS does not result in changes to procedures that could impact the 
occurrence of an accident.
    With respect to the issue of consequences of an accident, the 
function of the containment is to mitigate the radiological 
consequences of a loss of coolant accident (LOCA) or other 
postulated events that could result in radiation being released from 
the fuel inside containment. While the proposed change does not 
change the plant design, it does permit alteration of the 
containment boundary for the IFTS penetration. Altering the 
containment boundary in this case (i.e., removing the blind flange) 
results in some IFTS components possibly being subjected to 
containment pressure in the event of a LOCA. However, the additional 
post-accident peak pressure load to be imposed upon the components 
in the IFTS if the blind flange is removed is a small fraction of 
their design capability. Therefore, they are considered an 
acceptable barrier to prevent uncontrolled release of post-accident 
fission products for this proposed change.
    The proposed change required examination of two potential 
leakage pathways. The larger

[[Page 4274]]

is the IFTS transfer tube, itself. The other, much smaller one, is a 
branch line used for draining the IFTS transfer tube during its 
operation. It is clear that the gate valve at the bottom of the 
transfer tube is always water sealed and maintained so by the 
submergence of the water in the transfer tube and in the fuel 
building spent fuel storage pool (the lower pool). The height of 
this water seal is greater than that necessary to prevent leakage 
from the bottom of the transfer tube during accidents that result in 
the calculated peak post-DBA [design basis accident] LOCA pressure, 
Pa. Furthermore, the hydraulically operated gate valve in 
the lower end of the tube will remain closed, and has pressure 
retaining capability greater than that of the containment structure 
itself. The potential leakage pathway from the drain piping which 
attaches to the transfer tube will be isolated if required, via 
administrative controls on the drain piping isolation valve. 
Additionally, the drain piping isolation valve will be added to the 
Primary Containment Leakage Rate Testing Program (Technical 
Specification 5.5.13) to ensure that leakage past this valve will be 
maintained consistent with the leakage rate assumptions of the 
accident analysis. Due to the test methodology, the portion of the 
large transfer tube piping outboard of the blind flange (the portion 
of the tube which becomes exposed to the containment atmosphere 
during the draining portion of the IFTS operation) will also be part 
of the leakage rate test boundary and will therefore also be tested. 
Therefore, no unidentified leakage will exist from the piping and 
components that are outboard of the blind flange, and the leakage 
rate assumptions of the accident analysis will be maintained. Note 
that the bottom gate valve in the IFTS transfer tube will remain 
closed for this test evolution.
    Therefore, the proposed change does not result in a significant 
increase in the probability of the consequences of previously 
evaluated accidents, provided the bottom gate valve remains closed 
during MODE 1, 2, or 3 operation.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous analyzed.
    The proposed change consists of the removal of a passive 
component which is not part of the primary reactor coolant pressure 
boundary nor involved in the operation or shutdown of the reactor. 
Being passive, its presence or absence does not affect any of the 
parameters or conditions that could contribute to the initiation of 
any incidents or accidents that are created from a loss of coolant 
or an insertion of positive reactivity. Realigning the boundary of 
the primary containment to include portions of the IFTS is also 
passive in nature and therefore has no influence on, nor does it 
contribute to the possibility of a new or different kind of 
incident, accident or malfunction from those previously analyzed. 
Furthermore, operation of the IFTS is unrelated to the operation of 
the reactor and there is no mishap in the process that can lead to 
or contribute to the possibility of losing any coolant from the 
reactor or introducing the chance for an insertion of positive or 
negative reactivity, or any other accidents different from and not 
bounded by those previously evaluated.
    Therefore, the proposed change does not result in creating the 
possibility of a new or different kind of accident from any accident 
previously evaluated, provided the bottom gate valve remains closed 
during MODE 1, 2, or 3 operation.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed change involves the realignment of the primary 
containment boundary by removing the blind flange which is a passive 
component. The margin of safety that has the potential of being 
impacted by the proposed change involves the dose consequences of 
postulated accidents which are directly related to potential leakage 
through the primary containment boundary. The potential leakage 
pathways due to the proposed change have been reviewed, and leakage 
can only occur from the administratively controlled IFTS transfer 
tube drain piping, and from the IFTS transfer tube itself. A 
dedicated individual will be designated to provide timely isolation 
of this drain piping during the duration of time when this proposed 
change is in effect. The conservatively calculated dose which might 
be received by the designated individual while isolating the drain 
piping is calculated to be 3.8 rem TEDE [total effective dose 
equivalent], which remains within the guidelines of General Design 
Criterion (GDC) 19 (10 CFR 50, Appendix A, Criterion 19). 
Furthermore, the drain piping isolation valve will be added to the 
Primary Containment Leakage Rate Testing Program (Technical 
Specification 5.5.13) to ensure that leakage from the piping and 
components located outboard of the blind flange will be maintained 
consistent with the leakage rate assumptions of the accident 
analysis.
    Studies of the capability of the IFTS system to withstand 
containment pressurization under severe accident conditions have 
been conducted. These studies conclude that IFTS, including the 
transfer tube and its valves, has a capability to withstand beyond 
design basis severe accident containment pressures which is greater 
than that of the containment structure itself. The RBS Emergency 
Operating Procedures (EOPs) are based on an ultimate containment 
failure pressure capability of 53 psig [pounds per square inch--
gauge], which represents a margin of safety of 38 psi above the 15 
psig containment design pressure. This margin of safety is not 
impacted with the IFTS blind flange removed as long as the IFTS 
bottom valve remains closed. This capability to withstand 
containment pressurization under severe accident conditions envelops 
other non-DBA LOCA scenarios, such as the small break LOCA. For the 
large break LOCA, additional defense-in-depth is provided by 
maintaining a water seal greater than Pa above the outlet 
of the IFTS transfer tube in the lower pool.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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 Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: July 15, 1999 (NPF-38-216).
    Description of amendment request: One proposed change adds a 
Technical Specification (TS) Bases Control Program to the Waterford 3 
TS Administrative Controls Section, modeled after the guidelines 
contained in NUREG-1432. Additionally, the proposed change corrects an 
editorial error identified in the TS following issuance of Amendment 
146, dated October 19, 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. 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 changes to the Waterford 3 [Waterford 
Steam Electric Station, Unit 3] Technical Specifications add a TS 
Bases Control Program and correctly reference the appropriate 
document where administrative controls were relocated. The TS Bases 
Control Program will provide administrative controls that ensure 
changes to the TS Bases are appropriately reviewed and consistent 
with the Updated Final Safety Analysis Report (UFSAR). The addition 
of the proposed program does not affect any accident initiator or 
mitigation of any events analyzed in Chapter 15 of the UFSAR. Also, 
neither change has any affect on the operation of any structures, 
systems, or components or the assumptions of any accident analyses.
    The TS Bases Control Program will ensure that any change to the 
Bases that involves an unreviewed safety question will receive prior 
Nuclear Regulatory Commission approval. Changing the reference to 
the Quality Assurance Program Manual (QAPM) for the item relocated 
to the QAPM is purely administrative.

[[Page 4275]]

    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: The proposed changes to the Waterford 3 TS add a TS 
Bases Control Program and correctly reference the appropriate 
document where administrative controls were relocated. The addition 
of a TS Bases Control Program represents an administrative function 
performed under existing regulatory controls consistent with 10 CFR 
50.59. The proposed change to reference the appropriate document 
where an administrative control was relocated is purely 
administrative in nature. The change merely corrects the Technical 
Specifications wording to reflect the actual location of the record 
retention requirements for records of reviews performed on changes 
to the Process Control Plan (PCP) and Offsite Dose Calculation 
Manual (ODCM) in the QAPM.
    These proposed changes do not involve a change in plant design 
or affect the configuration or operation of any structure, system, 
or component, nor does it involve any potential initiating events 
that would create any new or different kind of accident. Therefore, 
the proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed changes to the Waterford 3 TS add a TS 
Bases Control Program and correctly reference the appropriate 
document where administrative controls were relocated. The addition 
of a TS Bases Control Program is an administrative change and has no 
[a]ffect on a margin of safety, as defined by Section 2 of the TS. 
The only [a]ffect of the TS Bases Control Program is to establish 
controls over how TS Bases changes are reviewed and implemented 
consistent with 10 CFR 50.59.
    The proposed change to a reference in the Administrative 
Controls section merely corrects the TS wording to reflect the 
actual location of the record retention requirements for records of 
reviews performed on changes to the PCP and ODCM in the QAPM.
    These proposed changes do not involve a change in plant design 
or have any affect on the plant protective barriers. Therefore, the 
proposed changes will not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: July 15, 1999 (NPF-38-217).
    Description of amendment request: The proposed change creates a 
new Technical Specification (TS) for the Main Feedwater Isolation 
Valves Section modeled after the guidelines of TS 3.7.3 in NUREG-
1432. Additionally, the letter provides for Nuclear Regulatory 
Commission (NRC) Staff review of an unreviewed safety question 
regarding the crediting of the Reactor Trip Override feature and 
Auxiliary Feedwater Pump high discharge pressure trip as assisting 
the operation of the Main Feedwater Isolation Valves during their 
required safety function, to close on a Main Steam Isolation Signal.
    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 to add the Main Feedwater 
Isolation Valves (MFIVs) to the Technical Specifications (TS) and 
provide an allowed outage time of 72 hours with appropriate required 
ACTIONs does not affect the operation of any structures, systems, or 
components or the assumptions of any accident analyses. The MFIVs 
are primarily designed to mitigate the consequences of a Main Steam 
Line Break (MSLB), and the Feedwater Line Break (FWLB). This TS 
change ensures the 5 second closure time currently assumed in the 
Waterford 3 [Waterford Steam Electric Station, Unit 3] analysis, 
thus it preserves the current analysis. Hence, the consequences of 
accidents previously evaluated do not change. Therefore, this change 
does not involve an increase in the consequences of any accident 
previously evaluated. Adding the MFIVs to the TS will not initiate 
an accident. Providing a TS and allowed outage time makes no changes 
to the plant and, thus, no increase in the probability of any 
accident previously evaluated.
    The accidents/events that may be affected by the proposed 
resolution to credit the Reactor Trip Override (RTO) circuitry for 
the Steam Generator [SG] Feed Pumps (SGFPs) during SGFP operation 
and the crediting of the Auxiliary Feedwater (AFW) pump high 
discharge pressure trip during AFW pump operation are the MSLB and 
the FWLB.
    The crediting of the RTO circuitry for the SGFPs and the 
crediting of the AFW pump trip will not affect the probability of 
occurrence of a MSLB or FWLB. Neither the SGFPs nor the AFW pump are 
initiators of either line break.
    The crediting of the RTO circuitry for the SGFPs and the 
crediting of the AFW pump trip will not adversely affect the 
consequences of a MSLB or FWLB. Ultimately, the RTO feature allows 
more reliable MFIV closure by reducing the differential pressure 
against which the MFIVs must close while not introducing a new 
failure mechanism such as a Loss of Feedwater or water hammer event.
    The RTO feature (which has always been a part of the Waterford 3 
plant design) mitigates the consequences of the MSLB and MFLB by 
reducing flow to the affected steam generator and containment.
    The Loss of Feedwater Event can be initiated by the loss of a 
SGFP. The currently analyzed Loss of Feedwater Event evaluates the 
loss of both SGFPs, which bounds a potential loss of one SGFP. 
Therefore, any modification that could increase the probability of a 
pump trip could increase the probability of this event. Since the 
proposed solution of crediting RTO features of the SGFPs and the 
trip of the AFW pump for the MFIV margin issue uses existing 
functions, no new features/trips will be added, and there is no 
increase in the probability or consequences of a Loss of Feedwater 
Event. The only plant modification being made is to enhance RTO such 
that it will run the SGFPs back to a minimum speed on a reactor 
trip, even when the FWCS [Feedwater Control System] is in manual. 
Although this slows the pump down, feedwater and the SGFPs remain 
available and the Loss of Feedwater Event probability is not 
significantly increased. The modification to make RTO function when 
the FWCS is in manual is not significant since the FWCS is in manual 
such a short period of time during plant operation.
    The AFW system is not credited in any accident analysis. The 
Emergency Feedwater (EFW) system is relied upon in the safety 
analyses to replenish SG inventory. Therefore, crediting the AFW 
pump discharge pressure trip will not involve an increase in the 
probability or consequences of any accident.
    In conclusion, the proposed TS change and resolution to the MFIV 
margin issue will not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of

[[Page 4276]]

accident from any accident previously evaluated?
    Response: The proposed TS change in itself does not change the 
design or configuration of the plant. No new or different equipment is 
being installed by the TS. No new or different accidents result from 
the addition of the MFIVs to the TS. Previously performed accident 
analyses remain valid. The proposed allowed outage time and required 
actions of the proposed TS do not change the procedural operation of 
the plant, but specify the requirements for treatment of the MFIVs 
under the plant TS. Therefore, no new or different type of accident 
from any accident previously evaluated is created.
    No new system interaction is created by crediting the existing RTO 
and AFW pump trip. Failure to isolate feedwater would require two 
failures, failure of the RTO or AFW circuitry, in addition to the 
failure of the Main Feedwater Regulating Valves (MFRVs) and Startup 
Feedwater Regulating Valves (SFRVs) to close, and is beyond single 
failure criteria. If the RTO and AFW features were the single failure, 
then closure of the regulating valves would be credited for MSIS [Main 
Steam Isolation Signal] isolation since the regulating valves were 
designed to close against SGFP shutoff head.
    RTO and AFW pump trips would not be considered initiators of a MSLB 
or FWLB, but could be considered initiators of a Loss of Feedwater 
Event. However, this event is bounded by the analyzed Waterford 3 Loss 
of Feedwater Events. No new event is created. The only hardware change 
being made is the use of RTO for pump run back when the FWCS is in 
manual. The existing signal will be used and routed through the same 
methods as are currently installed, ensuring it will run the pump back 
appropriately. Therefore, no new system interactions or events are 
created.
    The new method of potential failure that has not previously been 
evaluated is in the fact that Waterford 3 would now be crediting a non-
safety related circuit for closure of the safety related MFIVs. Non-
safety features are not normally credited for the proper operation of a 
safety related component. However, in this case, for the valve to close 
in the 5 seconds assumed in safety analyses, the RTO and AFW pump trip 
will be credited. Because this is new, different and not a previously 
approved allowance, this resolution must be submitted for NRC Staff 
approval. Entergy believes this resolution is acceptable based on the 
high degree of reliability of these components.
    The system design, as discussed above, does not increase the 
potential for a Loss of Feedwater Event and current analyses bound all 
potential accident scenarios. Therefore, the proposed TS change and 
resolution to the MFIV margin issue will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in a margin of safety?
    Response: The MFIVs have no [a]ffect on a margin of safety as 
defined by Section 2 of the TS. Their only [a]ffect is response to the 
accidents described above, which will be enhanced by specifying an 
allowed outage time, action requirements and surveillance requirements 
in the TS. Therefore, no reduction in the margin of safety is involved 
with the addition of these valves to the TS.
    No new system interaction is created by the crediting of the RTO 
feature or the AFW pump trip, or the addition of RTO operation in 
manual.
    The proposed resolution does affect a part of a protective 
boundary, the MFIV, which serves to isolate the Main Feedwater system 
from portions of the system inside containment. However, it does not 
affect operation or function of the valve itself since no changes to 
the valve are being made. The proposal allows increased margin for 
valve closure; therefore, margins of safety are not affected. The valve 
will close within the time limits required by safety analyses and 
general design criteria.
    Therefore, the proposed TS change and resolution to the MFIV margin 
issue 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, Esquire, Winston & Strawn, 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: July 15, 1999 (NPF-38-218).
    Description of amendment request: The proposed changes extend the 
Reactor Coolant System Pressure Temperature Curves to 20 Effective Full 
Power Years.
    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 changes will not increase the probability 
or consequences of any accident previously evaluated since the 
proposed changes revise the pressure/temperature limits in 
accordance with 10 CFR 50, Appendix G, utilizing the latest NRC 
[Nuclear Regulatory Commission] guidelines in Regulatory Guide 1.99, 
Revision 2, relative to estimating neutron irradiation damage to the 
reactor vessel. The proposed changes also maintain the conservative 
limits with respect to the low temperature overprotection (LTOP) 
system and heatup and cooldown restrictions.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed changes will not create the possibility 
of a new or different kind of accident from any previously analyzed 
since they do not introduce new systems, failure modes, or other 
plant perturbations. The proposed changes revise the pressure/
temperature limits in accordance with 10 CFR 50, Appendix G, 
utilizing the latest NRC guidelines in Regulatory Guide 1.99, 
Revision 2, relative to estimating neutron irradiation damage to the 
reactor vessel.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed changes will not involve a significant 
reduction in the margin of safety since equal or more stringent 
pressure/temperature limitation requirements for reactor operation 
will be applied. The proposed changes were derived in accordance 
with approved NRC methodology which was developed to assure the 
reactor coolant system pressure boundary is designed with sufficient 
margin to withstand any condition during normal operation including 
anticipated operational occurrences and system inservice leak and 
hydrostatic tests.
    These requirements were revised in accordance with 10 CFR 50, 
Appendix G, utilizing the latest NRC guidance in Regulatory Guide 
1.99, Revision 2, relative to estimating neutron irradiation damage 
to the reactor vessel. The LTOP system limits were also reanalyzed 
for the proposed changes.

[[Page 4277]]

    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, Esquire, Winston & Strawn, 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: July 19, 1999. (NPF-38-219).
    Description of amendment request: The proposed changes modify 
Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical 
Specification (TS) 4.5.2.f.2 by increasing the performance requirement 
for the low pressure safety injection (LPSI) pumps. The change revises 
the LPSI pump Surveillance Requirements to measure pump developed head, 
instead of pump discharge pressure. The associated changes to TS Bases 
are included in the submittal.
    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: Increasing the LPSI pump performance requirements will 
not increase the probability or consequences of any accidents. There 
are no physical changes to the pump. The only procedure changes 
required are to Surveillance Procedure OP-903-030, ``Safety 
Injection Pump Operability Evaluation.'' The changes do not impact 
plant operating procedures. The LPSI system is primarily designed to 
mitigate the consequences of a large break Loss of Coolant Accident 
(LOCA). These proposed changes do not affect any of the assumptions 
used in the deterministic LOCA analysis. Hence the consequences of 
accidents previously evaluated do not change.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed change does not alter plant operations, 
nor does it alter the physical plant. The change only increases 
existing equipment performance requirements. No different accidents 
result from the increase in performance requirements. No change is 
being made to the parameters within which the plant is operated. The 
setpoints at which protective or mitigative actions are initiated 
are unaffected by this change. 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. The proposed change will only increase the performance 
requirements of the LPSI pumps.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: To the contrary, the change increases LPSI pump 
performance requirements, increasing the margin between the TS 
performance requirements and the analytical limit.
    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, Esquire, Winston & Strawn, 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: August 4, 1999 (NPF-38-222).
    Description of amendment request: The proposed change modifies 
Technical Specifications (TS) 3.5.2 to extend the allowed outage time 
(AOT) to seven days for one high pressure safety injection (HPSI) train 
inoperable and TS 3.5.3 to change the end-state to HOT SHUTDOWN with at 
least one OPERABLE shutdown cooling train in operation. Additionally, 
an AOT of 72 hours in TS 3.5.2 is imposed for other conditions where 
the equivalent of 100 percent emergency core cooling system (ECCS) 
subsystem flow is available. If 100 percent ECCS flow is unavailable 
due to two inoperable HPSI trains, an ACTION has been added to restore 
at least one HPSI to OPERABLE status within one hour or place the plant 
in HOT STANDBY in six hours and to exit the MODE of applicability in 
the following six hours. In the event the equivalent of 100 percent 
ECCS subsystem flow is not available due to other conditions, TS 3.0.3 
is entered. The Limiting Condition for Operation terminology is being 
changed for consistency with the ECCS requirements. Additionally, the 
associated TS Bases are being changed.
    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 High Pressure Safety Injection System (HPSI) is 
part of the Emergency Core Cooling System subsystem. Inoperable HPSI 
components are not accident initiators in any accident previously 
evaluated. Therefore, this change does not involve an increase in 
the probability of any accident previously evaluated.
    The HPSI system is primarily designed to mitigate the 
consequences of a Loss of Coolant Accident (LOCA). These proposed 
changes do not affect any of the assumptions used in the 
deterministic LOCA analyses. Hence the consequences of accidents 
previously evaluated do not change.
    In order to fully evaluate the HPSI AOT extension, probabilistic 
safety assessment (PSA) methods were utilized. The results of these 
analyses show no significant increase in the core damage frequency. 
These analyses are detailed in report CE NPSD-1041, ``Joint 
Applications Report for High Pressure Safety Injection System 
Technical Specification Modifications,'' March 1998.
    The Configuration Risk Management Program is an Administrative 
Program that assesses risk based on plant status. Adding the 
requirement to implement this program for Technical Specification 
3.5.2 does not affect the probability or the consequences of an 
accident.
    The proposed change allows a combination of equipment from 
redundant trains to be inoperable provided that at least the 
equivalent of a single ECCS subsystem remains operable. Analyzed 
events are assumed to be initiated by the failure of plant 
structures, systems or components. Allowing equipment from redundant 
trains to constitute a single operable subsystem does not increase 
the probability that a failure leading to an analyzed event will 
occur. The ECCS components are passive until an actuation signal is 
generated. This change does not increase the failure probability of 
the ECCS components. This change reduces the plant's susceptibility 
to common cause failures. As such, the probability of occurrence for 
a previously analyzed accident are not significantly increased.

[[Page 4278]]

    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: The proposed change does not change the design or 
configuration of 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, and the setpoints at which protective or 
mitigative actions are initiated are unaffected by this change. 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. The proposed change 
will only provide the plant some flexibility in maintaining the 
minimum equipment required to be operable to perform the ECCS 
function while in this condition. The change does not alter 
assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed changes do not affect the limiting 
conditions for operation or their bases used in the deterministic 
analysis to establish the margin of safety. PSA evaluations were 
used to evaluate these changes. These evaluations demonstrate that 
the changes involve no significant increase in risk. These 
evaluations are detailed in report CE NPSD-1041. The margin of 
safety is established through equipment design, operating 
parameters, and the setpoints at which automatic actions are 
initiated. None of these are adversely impacted by the proposed 
change. Sufficient equipment remains available to actuate upon 
demand for the purpose of mitigating a transient event. The proposed 
change, which allows operation to continue for up to 72 hours with 
components inoperable in both ECCS subsystems, is acceptable based 
on the remaining ECCS components providing 100% of the required ECCS 
flow. The reduced potential for a self-induced plant transient 
resulting from unit shutdown required for a second inoperable ECCS 
train is minimized. Therefore, the change does not involve a 
significant reduction in the margin of safety, and is offset by 
minimizing the potential for a self induced plant transient.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: August 4, 1999 (NPF-38-223).
    Description of amendment request: The proposed change modifies 
Technical Specification (TS) 3.5.2 to extend the allowed outage time 
(AOT) to seven days for one low pressure safety injection (LPSI) train 
inoperable. Additionally, an AOT of 72 hours is imposed for other 
conditions where the equivalent of 100 percent emergency core cooling 
system (ECCS) subsystem flow is available. If 100 percent ECCS flow is 
unavailable due to two inoperable LPSI trains, an ACTION has been added 
to restore at least one LPSI train to OPERABLE status within one hour 
or place the plant in HOT STANDBY in six hours and to exit the MODE of 
applicability in the following six hours. In the event the equivalent 
of 100 percent ECCS subsystem flow is not available due to other 
conditions, TS 3.0.3 is entered. The Limiting Condition for Operation 
terminology is being changed for consistency with the ECCS 
requirements. Additionally, the associated TS Bases are being changed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No. The Low Pressure Safety Injection System (LPSI) is 
part of the Emergency Core Cooling System subsystem. Inoperable LPSI 
components are not accident initiators in any accident previously 
evaluated. Therefore, this change does not involve an increase in 
the probability of an accident previously evaluated.
    The LPSI system is primarily designed to mitigate the 
consequences of a large Loss of Coolant Accident (LOCA). These 
proposed changes do not affect any of the assumptions used in the 
deterministic LOCA analysis. Hence, the consequences of accidents 
previously evaluated do not change.
    In order to fully evaluate the LPSI AOT extension, probabilistic 
safety analysis (PSA) methods were utilized. The results of these 
analyses show no significant increase in the core damage frequency. 
As a result, there would be no significant increase in the 
consequences of an accident previously evaluated. These analyses are 
detailed in CE NPSD-995, Combustion Engineering Owners Group ``Joint 
Applications Report for Low Pressure Safety Injection System AOT 
Extension.''
    The Configuration Risk Management Program is an Administrative 
Program that assesses risk based on plant status. Adding the 
requirement to implement this program for Technical Specification 
3.5.2 does not affect the probability or the consequences of an 
accident.
    The proposed change allows a combination of equipment from 
redundant trains to be inoperable provided that at least the 
equivalent of single train of ECCS remains operable. Analyzed events 
are assumed to be initiated by the failure of plant structures, 
systems or components. Allowing equipment from redundant trains to 
constitute a single operable train does not increase the probability 
that a failure leading to an analyzed event will occur. The ECCS 
components are passive until an actuation signal is generated. This 
change does not increase the failure probability of the ECCS 
components. This change reduces the plant's susceptibility to common 
cause failures. As such, the probability of occurrence for a 
previously analyzed accident are not significantly increased.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No. The proposed change does not change the design or 
configuration of 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, and the setpoints at which protective or 
mitigative actions are initiated are unaffected by this change. 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. The proposed change 
will only provide the plant some flexibility in maintaining the 
minimum equipment required to be operable to perform the ECCS 
function while in this Condition. The change does not alter 
assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

[[Page 4279]]

    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No. The proposed changes do not affect the limiting 
conditions for operation or their bases used in the deterministic 
analyses to establish the margin of safety. PSA evaluations were 
used to evaluate these changes. These evaluations demonstrate that 
the changes are either risk neutral or risk beneficial. These 
evaluations are detailed in CE NPSD-995. The margin of safety is 
established through equipment design, operating parameters, and the 
setpoints at which automatic actions are initiated. None of these 
are adversely impacted by the proposed change. Sufficient equipment 
remains available to actuate upon demand for the purpose of 
mitigating a transient event. The proposed change, which allows 
operation to continue for up to 72 hours with components inoperable 
in both ECCS trains, is acceptable based on the remaining ECCS 
components providing 100% of the required ECCS flow. The reduced 
potential for a self-induced plant transient resulting from unit 
shutdown required for a second inoperable ECCS train is minimized. 
Therefore, the change does not involve a significant reduction in 
the margin of safety, and is offset by minimizing the potential for 
a self induced plant transient.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

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

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: December 3, 1998.
    Description of amendment requests: The proposed amendments would 
add a new Technical Specification (T/S) and associated Bases for the 
distributed ignition system (DIS). The proposed change incorporates the 
technical requirements of NUREG-1431, Revision 1, ``Standard Technical 
Specifications, Westinghouse Plants.''
    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 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The T/S being proposed for the DIS is consistent with its design 
and operation as previously reviewed and approved, and therefore, 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The amendments 
involve new requirements for the T/Ss and do not delete any existing 
requirements.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident previously evaluated.
    The T/S being proposed for the DIS is consistent with its design 
and operation as previously reviewed and approved, and therefore, 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The T/S being proposed for the DIS is consistent with [the] 
design and operation as previously reviewed and approved, and 
therefore, does not involve a significant reduction in a margin of 
safety. Compliance with the proposed T/S will provide additional 
assurance of system availability to maintain a margin of safety for 
containment integrity during degraded core events.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: December 15, 1999.
    Description of amendment request: This proposed technical 
specification (TS) change will revise the average power range Monitors 
(APRMs) neutron flux-high (flow biased) allowable value based on a 
revised power to flow map. The revised power to flow map extends the 
current plant operating domain to above the rated rod line, to within 
an envelope referred to as the maximum extended load line limit (MELLL) 
and adds the increased core flow (105%) region. The current power to 
flow map is based on a region bounded by the extended load line limit 
(ELLL) and evaluations prepared as part of the Core Operating Limits 
Report (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. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Attachment 3 [to the December 15,1999 
application] (Reference 1) evaluates operation in the Maximum 
Extended Load Line Limit (MELLL) and Increased Core Flow (ICF) 
regions and the impact on equipment and safety system performance. 
Impacts on containment, the reactor vessel, Recirculation System, 
reactor vessel internals, limiting transients for the Cycle 20 
reload (upcoming refuel outage), Loss of Coolant Accident (LOCA), 
and Anticipated Transients Without SCRAM (ATWS) events were 
evaluated. The conclusion is that for all events, accidents, and 
equipment evaluated, operation and event response remain within 
previously established design limits and acceptance criteria. No 
changes in the initiators of accidents previously evaluated are 
being made by this change. Because operation in the expanded regions 
maintains adequate design margin and there are no changes in the 
accident initiators, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    In support of operation in the MELLL region, the proposed change 
modifies (increases) the Average Power Range Monitor (APRM) Neutron 
Flux-High (Flow Biased) allowable value. Changes to the setpoint and 
allowable value will be implemented in accordance with approved 
setpoint methodology and plant procedures (References 7 and 8). As 
noted in Technical Specifications (TS) Bases Section B.3.3.1.1.2.b: 
``No specific safety analyses take credit for the APRM Neutron Flux-
High (Flow Biased) Function.'' The APRM allowable value credited in 
accident analyses is based on the 120% fixed scram-allowable value 
(TS Table 3.3.1.1-1, Function 2.c), which remains unchanged as a 
result of this requested TS change. Though not credited in analyses, 
the limiting flow biased value of 119% Reactor Thermal Power (RTP) 
also remains unchanged. Evaluations presented in Attachment 3 
demonstrate that operation in the MELLL envelope, with reliance on 
the credited fixed scram allowable value (analytically assumed at 
123% RTP to justify a 120% TS allowable value), results in event and 
accident responses within design limits and established acceptance 
criteria. Therefore, no significant increase in source term, 
radiological consequences or other accident consequences occurs as a 
result of the proposed change.
    The proposed change has no affect on operation in the ICF 
region. The allowable value, as part of the proposed change, will 
reach its clamped upper limit value of 119% reactor thermal power. 
Core flows at or above this level will result in the allowable value 
reaching its current TS upper limit of 119%. As stated above, the 
limiting value remains unchanged as part of this request.
    The postulated failure mechanisms for the equipment are not 
changed, nor are any

[[Page 4280]]

design limits exceeded. The proposed change will result in the need 
to replace APRM equipment to allow operation in the extended power 
to flow domain. These replacements will be evaluated per the 
requirements of 10 CFR 50.59 as part of the Cooper Nuclear Station 
(CNS) design change process to confirm no Unreviewed Safety Question 
is created. Therefore, implementation of this proposed TS amendment 
will not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change will not create the possibility of a new 
or different kind of accident than previously evaluated.
    This proposed change does not modify the functional requirements 
of the affected equipment, create any new system interfaces or 
interactions, create any new process conditions that exceed design 
limits, nor create any new system failure modes or sequences of 
events that could lead to an accident.
    The postulated failure mechanisms for the equipment are not 
changed, nor are any design limits or acceptance criteria exceeded. 
The proposed change will result in the need to replace APRM 
equipment to allow operation in the extended power to flow domain. 
These replacements will be evaluated per the requirements of 10 CFR 
50.59 as part of the CNS design change process to confirm no 
Unreviewed Safety Question is created. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change will not involve a significant reduction 
in a margin of safety.
    Change to the APRM Neutron Flux-High (Flow Biased) allowable 
value is still limited by the 119% RTP value of TS. This value is 
not credited in the safety analyses. In addition, the existing 120% 
fixed scram allowable value (TS Table 3.3.1.1-1, Function 2.c) still 
provides the same margin to the Analytical Limit of 123% RTP. 
Analyses documented in Attachment 3 demonstrate that for operation 
in the MELLL envelope or ICF region, adequate margin to design 
limits is maintained and event acceptance criteria are met. Thus, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: December 22, 1999.
    Description of amendment request: The proposed license amendment 
requests Nuclear Regulatory Commission (NRC) review and approval of 
revisions to the Cooper Nuclear Station (CNS) design basis accident 
(DBA) radiological assessment calculational methodology used to 
demonstrate compliance with the Exclusion Area Boundary and Low 
Population Zone dose acceptance criteria specified in 10 CFR 100.11, 
and the control room dose acceptance criteria discussed in General 
Design Criteria (GDC) 19 of 10 CFR 50, Appendix A. The revisions entail 
a complete rewrite of the radiological assessment calculational 
methodology. The proposed changes do not revise the accident category, 
general accident description, identification of accident cause, 
frequency classification, starting conditions of the accident, accident 
sequence of events, or system operation as described in the CNS Updated 
Safety Analysis Report (USAR). The revised radiological assessment 
calculational methodology does, however, involve changes to the 
radiological consequence summary, fission product release from fuel 
assumptions, fission product release to secondary containment 
assumptions and conditions, fission product release to the environs 
assumptions and initial conditions, and radiological effects summary 
described in the CNS USAR. Additionally, the revised CNS DBA 
radiological assessment calculational methodology incorporates the GDC 
19 control room dose acceptance criteria determination as part of the 
assessment. Previously the control room dose assessment was maintained 
as separate design calculations and not included in the CNS USAR DBA 
radiological assessment summaries.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revisions to the Design Basis Accident (DBA) 
radiological assessment calculational methodology do not affect the 
accident initiators or precursors of accidents previously evaluated. 
The proposed revisions to the methodology do not affect the existing 
design, function or operation of systems, structures or components 
in the facility. No new or different type of plant equipment is 
installed by the revised radiological assessment calculational 
methodology. Plant operating modes are not changed due to the 
proposed revision to the DBA radiological assessment calculational 
methodology. The proposed revisions are calculational in nature and 
serve only to incorporate more recent site specific meteorological 
data, reflect plant specific system operating parameters and design, 
utilize more widely accepted accident assumptions for a facility of 
Cooper Nuclear Station's vintage, incorporate the Technical 
Information Document (TID-14844) source term to be consistent with 
the accident assumptions used, update fuel parameter considerations 
to include higher burnup fuel designs, and to utilize generic and 
updated calculational and software methodologies to perform the 
analysis. These revisions improve the consistency between the 
accident dose calculation assumptions and improve the documentation 
basis for each accident calculation. The revisions utilize 
conservatively lower accident mitigation system filter efficiency 
assumptions and incorporate plant specific accident mitigation 
system operating parameter and design assumptions which result in a 
calculated radiological consequence increase. Operation of accident 
mitigation systems, structures and components is not altered by the 
changes in accident mitigation assumptions. Due to the broad changes 
in the calculational methodology and assumptions, and an increase in 
the postulated accident source term, the calculated radiological 
dose consequences of each design basis accident have changed and in 
some cases increased. In each case, however, the calculated 
radiological dose consequences satisfy the Exclusion Area Boundary 
and Low Population Zone radiological dose acceptance criteria 
specified in 10 CFR 100 and the control room dose acceptance 
criteria discussed in General Design Criteria 19 (GDC 19) of 10 CFR 
50, Appendix A. Therefore, the proposed revisions do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does not create the possibility for a new or different kind 
of accident from any accident previously evaluated.
    The proposed revisions to the DBA radiological assessment 
calculational methodology do not change the existing design, 
function or operation of systems, structures or components in the 
facility. No new or different type of plant equipment is installed 
by this change. There are no changes to existing design parameters 
governing plant operation, plant operating modes, or changes in 
system interfaces. No new types of accident initiators or precursors 
are created by the proposed revision to the DBA radiological 
assessment calculational methodology. The proposed revisions are 
calculational in nature and serve only to incorporate more recent 
site specific meteorological data, reflect plant specific system 
operating parameters and design, utilize more widely accepted 
accident assumptions for a facility of Cooper Nuclear Station's 
vintage, incorporate the TID-14844 source term to be consistent with 
the accident assumptions used, update fuel parameter considerations 
to include higher burnup fuel designs, and to utilize generic and 
updated calculational and software

[[Page 4281]]

methodologies to perform the analysis. These revisions improve the 
consistency between the accident dose calculation assumptions and 
improve the documentation basis for each accident calculation. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident previously evaluated.
    3. Does not create a significant reduction in the margin of 
safety.
    The proposed revisions to the DBA radiological assessment 
calculational methodology do not involve a relaxation in the 
criteria used to establish safety limits or a relaxation in the 
limiting conditions for operation. The accident analysis sequence of 
events remains unchanged. The proposed change will not result in any 
challenges to plant equipment, fuel integrity, or the reactor 
coolant system pressure boundary. The proposed revisions are 
calculational in nature and serve only to incorporate more recent 
site specific meteorological data, reflect plant specific system 
operating parameters and design, utilize more widely accepted 
accident assumptions for a facility of Cooper Nuclear Station's 
vintage, incorporate the TID-14844 source term to be consistent with 
the accident assumptions used, update fuel parameter considerations 
to include higher burnup fuel designs, and to utilize generic and 
updated calculational and software methodologies to perform the 
analysis. These revisions improve the consistency between the 
accident dose calculation assumptions and improve the documentation 
basis for each accident calculation. Therefore, the proposed change 
does not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 19, 1999.
    Description of amendment request: The licensee proposes to change 
the Technical Specifications (TS) by relocating the specific 
requirements of TS 6.4.3, ``Nuclear Safety Audit Review Committee 
(NSARC),'' to the Quality Assurance Program located in the Updated 
Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. The proposed change does not alter or prevent the 
ability of structures, systems, or components (SSCs) to perform 
their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR). The proposed change is 
administrative in nature and does not decrease the effectiveness of 
programmatic controls or the procedural details of assuring 
operation of the facility in a safe manner.
    The relocation of the Nuclear Safety Audit Review Committee 
requirements from the Technical Specification to a new Appendix 17C 
in UFSAR Chapter 17.2 does not alter the performance or frequency of 
these activities. Future changes to the Quality Assurance Program 
are subject to the 10 CFR 50.54(a) and 10 CFR 50.59 and change 
processes.
    The proposed change will not degrade the ability of systems, 
structures and components important to safety to perform their 
safety function. The proposed change will not change the response of 
any system, structure or component important to safety as described 
in the UFSAR. Since the plant response to an accident will not 
change, there is no change in the potential for an increase in the 
consequences of an accident previously analyzed. As such, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed change does not alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. There are no changes to the source term, 
containment isolation or radiological release assumptions used in 
evaluating the radiological consequences in the Seabrook Station 
UFSAR. Existing system and component redundancy is not being changed 
by the proposed change. The proposed change has no adverse impact on 
component or system interactions. The proposed change will not 
adversely degrade the ability of systems, structures and components 
important to safety to perform their safety function nor change the 
response of any system, structure or component important to safety 
as described in the UFSAR. The proposed change is administrative in 
nature and does not change the level of programmatic controls and 
procedural details of assuring operation of the facility in a safe 
manner. The proposed changes involve the relocation of the 
requirements of the Nuclear Safety Audit Review Committee from TS 
6.4.3 to Updated Final Safety Analysis Report, Chapter 17.2, 
``Quality Assurance Program'' in a new Appendix 17C. Future changes 
to the Quality Assurance Program are subject to the 10 CFR 50.54(a) 
and 10 CFR 50.59 and change processes.
    Therefore, since there are no changes to the design assumptions, 
conditions, configuration of the facility, or the manner in which 
the plant is operated and surveilled, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes involve the relocation of the requirements 
of the Nuclear Safety Audit Review Committee from TS 6.4.3 to 
Updated Final Safety Analysis Report, Chapter 17.2, ``Quality 
Assurance Program'' in a new Appendix 17C. There is no adverse 
impact on equipment design or operation and there are no changes 
being made to the Technical Specification required safety limits or 
safety system settings that would adversely affect plant safety. The 
proposed change is administrative in nature and does not change the 
level of programmatic controls and procedural details controls of 
assuring operation of the facility in a safe manner.
    Future changes to the Quality Assurance Program are subject to 
the 10 CFR 50.54(a) and 10 CFR 50.59 change processes. Therefore, 
relocation of the requirements contained in TS 6.4.3 to the Update 
Final Safety Analysis Report does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: November 29, 1999.
    Description of amendment request: The licensee proposes to change 
Technical Specification (TS) Surveillance Requirement (SR) 4.8.1.1.2f., 
to relocate sub requirement 4.8.1.1.2f.1 which requires inspection of 
the emergency diesel generators (EDGs) on an 18-month cycle to be 
subjected to an inspection in accordance with manufacturers 
recommendations, to the Seabrook Station Technical Requirements Manual 
(SSTRM).
    Basis for proposed no significant hazards consideration 
determination:

[[Page 4282]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated. The proposed change does not alter or prevent 
the ability of structures, systems and components (SSCs) to perform 
their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR).
    Performance of EDG inspection activities based on condition-
based maintenance rather than time-directed maintenance will neither 
exacerbate nor significantly increase the probability or 
consequences of an accident previously evaluated in the Seabrook 
Station UFSAR. North Atlantic has extensive experience and expertise 
in operating and maintaining the EDGs to determine the appropriate 
maintenance activities for demonstrating operability of the EDGs. 
North Atlantic will continue to use, in conjunction with 
manufacturer recommendations, prudent engineering judgment when 
conducting testing, preventive and corrective maintenance activities 
on the EDGs. In addition, the other surveillance testing required by 
SR 4.8.1.1.2f would continue to ensure that the EDGs are capable of 
performing their safety function.
    Throughout the first six fuel cycles, overall EDG condition has 
steadily improved with the use of improved design, utilization of 
better condition monitoring tools and procedures and the reduction 
of intrusive preventative maintenance tasks made possible by the 
improved on-line condition monitoring methods. These improvements 
resolved problems that were recognized during the early years of EDG 
operation.
    North Atlantic has implemented the Maintenance Rule Program in 
accordance with the provisions of 10 CFR 50.65, Regulatory Guide 
(RG) 1.160, and NUMARC 93-01, ``Industry Guide for Monitoring the 
Effectiveness of Maintenance at Nuclear Power Plants.''
    North Atlantic's maintenance rule program establishes specific 
performance criteria for SSCs. Reliability and unavailability 
performance criteria have been assigned to risk significant and 
standby safety-related non-risk significant SSCs. Other in-scope 
SSCs have been assigned appropriate reliability and/or plant level 
performance criteria. SSCs that are determined to not meet the 
established performance criteria are designated as (a)(1) and are 
subject to action plans, goal setting, and goal monitoring. 
Performance of (a)(1) SSCs is compared to the established goals. 
When it is determined that the performance goals have been achieved, 
a SSC may be returned to the normal performance monitoring (a)(2) 
status.
    With regard to the EDGs, these components and the associated 
support systems are risk significant and standby safety-related. The 
experience to date, applying the Maintenance Rule Program to the 
EDGs, has proven to be positive. Risk informed decision-making 
concerning the benefits of maintenance and time out of service has 
maintained reliable EDGs with unavailability consistent with the 
assumptions in the Seabrook Station Probabilistic Risk Assessment 
(PRA).
    Furthermore, Operations Department personnel perform daily, 
weekly, biweekly, monthly and quarterly walkdowns and inspections of 
various items as well as the monthly surveillance run on each 
diesel. These inspections, combined with system control panel 
alarms, engine oil sampling and on-line monitoring of engine 
vibration and running performance (cylinder firing, fuel delivery 
and exhaust temperatures), enable expeditious response to a 
developing degraded condition and provide a mechanism for failure 
identification prior to performance of the refueling interval 
surveillances.
    Based on the reviews of the surveillance tests, inspections and 
maintenance activities, it is concluded that there is no significant 
impact on the reliability of the EDGs and, therefore, there is no 
significant increase in the probability or consequences of any 
previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated. There are no changes to the source term, 
containment isolation or radiological release assumptions used in 
evaluating the radiological consequences in the Seabrook Station 
UFSAR. Existing system and component redundancy is not being changed 
by the proposed change. The proposed change has no adverse affect on 
component or system interactions. Therefore, since there are no 
changes to the design assumptions, conditions, configuration of the 
facility, or the manner in which the plant is operated, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not adversely affect equipment design 
or operation and there are no changes being made to the Technical 
Specification required safety limits or safety system settings that 
would adversely affect plant safety. The proposed change does not 
adversely affect the EDG's ability to ensure that sufficient power 
is available to supply the safety related equipment required for: 1) 
the safe shutdown of the facility, and 2) the mitigation and control 
of accident conditions within the facility.
    Surveillance testing of the EDGs during normal plant operation 
provides assurance that the proposed change will not adversely 
affect the reliability of the EDGs. North Atlantic will continue to 
use, in conjunction with manufacturer's recommendations, prudent 
engineering judgment when conducting testing, preventive, and 
corrective maintenance activities on the EDGs. In addition, the 
other surveillance testing required by SR 4.8.1.1.2f would continue 
to ensure that the EDGs are capable of performing their safety 
function. Thus, it is concluded that the EDGs would continue to be 
available upon demand to mitigate the consequences of an accident 
and, therefore, there is no significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities.Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: December 3, 1999.
    Description of amendment request: The licensee proposes to change 
the technical specifications (TS) by incorporating reference to the 
American Society for Testing and Materials (ASTM) Standard D3803-1989, 
``Standard Test Method for Nuclear-Grade Activated Charcoal,'' as the 
test protocol for charcoal filter laboratory testing. In addition, 
there will be a change to Surveillance Requirements 4.7.6.1d.5) and 
4.9.12d.4) specifying a minimum required heater output based on design 
rated voltage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect accident initiators or 
precursors and do not alter the design assumptions, conditions or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, or components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR).
    The proposed changes modify the Technical Specifications to 
reference

[[Page 4283]]

appropriate test parameters for performing laboratory testing of 
nuclear-grade charcoal in ESF [engineered safety feature] filtration 
systems in accordance with ASTM D3803-89. The testing methodology 
associated with ASTM D3803-89 provides more stringent requirements 
than what is currently employed. These more stringent requirements 
will not result in operations that will increase the probability of 
initiating an analyzed event and do not alter assumptions relative 
to mitigation of an accident or transient event. The more 
restrictive requirements continue to ensure process variables, 
structures, systems, and components are maintained consistent with 
the safety analyses and licensing basis.
    The proposed change associated with verification of heater 
capacity dissipation by specifying a minimum required output based 
on design rated voltage does not affect continued operability of the 
heater. Stipulating the design rated voltage ensures the heater(s) 
remains capable of performing its safety function. Specifying an 
upper kW range band is restrictive and has been determined to be 
unnecessary. There is no safety concern with the heaters operating 
at a higher kW output. Operating at a higher kW output improves 
dehumidification. Should maximum operating bus voltage conditions be 
experienced it does not pose a fire hazard or dry-out concern for 
the charcoal filters.
    There are no changes to previous accident analyses. The 
radiological consequences associated with these analyses remain 
unchanged. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed changes do not alter the design assumptions, 
conditions or configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes have no 
impact on component or system interactions.
    The proposed changes modify the Technical Specifications to 
reference appropriate test parameters for performing laboratory 
testing of nuclear-grade charcoal in ESF filtration systems in 
accordance with ASTM D3803-89. The changes do impose different, more 
conservative testing requirements, on the ESF filtration systems 
charcoal samples. However, there is no alteration in the methods 
employed to obtain the charcoal sample and testing is performed 
offsite.
    The proposed change associated with verification of heater 
capacity dissipation by specifying a minimum required output based 
on design rated voltage does not affect continued operability of the 
heater. The design function of the heater for humidity control 
remains unchanged. Deletion of the upper kW range does not pose a 
fire or dry-out concern for the charcoal filters.
    These changes are consistent with the safety analyses and 
licensing basis. The proposed changes do not introduce any new modes 
of plant operation, or alter any operational setpoints.
    Since the proposed changes do not involve the physical 
alteration of SSCs (i.e., no new or different type of equipment to 
be installed) or changes in the methods governing normal plant 
operation, it is concluded that the proposed changes do not create 
the possibility of a new or different kind of accident from any 
previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There is no impact on equipment design or operation and there 
are no changes being made to the Technical Specification required 
safety limits or safety system settings that would adversely affect 
plant safety. The proposed changes modify the Technical 
Specifications to reference appropriate test parameters for 
performing laboratory testing of nuclear-grade charcoal in ESF 
filtration systems in accordance with ASTM D3803-89. The imposition 
of the more conservative charcoal filter testing requirements 
associated with ASTM D3803-89 has no significant impact on a margin 
of safety. The conservative nature of ASTM D3803-89 is by 
definition, providing additional restrictions to enhance plant 
safety.
    The proposed change associated with specifying a minimum 
required heater output based on design rated voltage does not reduce 
the ability of the heater to provide the minimum required kW output 
for humidity control. Deletion of the upper kW range does not pose a 
fire or dry-out concern for the charcoal filters.
    The proposed changes maintain requirements within the safety 
analysis and licensing basis. Therefore, the proposed changes do not 
involve a significant reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: September 7, 1999.
    Description of amendment request: The proposed changes affect 
Technical Specification 3/4.7.8, ``Plant Systems, Snubbers,'' by 
removing the current special exception which precludes applying the 
eighteen month functional testing surveillance to the Steam Generator 
Hydraulic Snubbers.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The snubbers provide a restraint function to mitigate the 
consequences of a Main Steam Line Break (MSLB) or to limit seismic 
induced movements of the steam generators so as to protect the 
attached Reactor Coolant System (RCS) piping and therefore prevent 
the initiation of a Loss of Coolant Accident (LOCA).
    While the proposed surveillance changes will extend the time 
period required for 100% inspection of all steam generator snubbers 
and also the actual service life of the snubber seals, the testing 
of samples at reduced intervals will actually provide a more 
reliable and timely indication of snubber functionality and provide 
increased assurance that generic concerns associated with this 
snubber set will be detected prior to any failure. The proposed 
surveillance requirements are the same as currently used for the 
balance of Millstone Unit No. 2 hydraulic snubbers. Given the 
complete similarity of design and operation for these components, 
the sampling approach is well suited for these snubbers. Given the 
general acceptance of a 10% sampling approach in the general snubber 
population, its use here for this homogenous set of components is 
fully justified. In addition to the 10% sample that will be 
functionally tested on an eighteen month interval, a concurrent 100% 
visual inspection is conducted during each test period, providing 
added assurance that no seal failures will go undetected for any 
significant period. This visual inspection program is unchanged from 
the existing surveillance program as currently documented in the 
Millstone Unit No. 2 Technical Specification. The anticipated 
reliability under the new surveillance frequency and testing methods 
proposed for the steam generator snubbers will not affect the 
probability of occurrence of a LOCA or a MSLB as the snubbers' 
ability to perform their function will prevent over stressing of 
either the Main Steam (MS) or RCS piping attached to the steam 
generators. Furthermore, the anticipated reliability under the new 
surveillance frequency and testing methods proposed for the steam 
generator snubbers will ensure that the existing evaluated 
consequences for these accidents will not be increased. Therefore, 
these changes will not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to Bases Section 3/4.7.8 will delete the 
text associated with the current exception taken for steam generator 
snubbers. This change will make the discussion in the Bases 
consistent with the proposed Technical Specification changes. 
Therefore, this change will not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not alter how any structure, system, or 
component functions. There will be no effect on equipment important 
to safety. The proposed changes have no effect on any of the design 
basis accidents previously evaluated. Therefore, this License 
Amendment Request does not impact the probability of an

[[Page 4284]]

accident previously evaluated, nor does it involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The only accidents possible due to failure of the steam 
generator snubbers to operate properly is increased stresses on both 
the MS and RCS piping attached to the steam generator due to either 
additional constraint in the case of premature lockup, or lack of 
proper constraint in the case of failure to lock-up under dynamic 
loading. Since the worst case scenario of such a failure would be 
the initiation of a LOCA, which is currently evaluated in the SAR 
[safety analysis report], there is no possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will allow use of the preferred approach to 
snubber surveillance which is in effect for the balance of Millstone 
Unit No. 2 snubbers. The steam generator snubbers have been 
previously exempt from the standard approach to snubber surveillance 
due to the difficulty previously encountered in testing these large 
and inaccessible components. Given the reliability of these snubbers 
is not expected to change in that the same requirements as for all 
other hydraulic snubbers will now consistently be met, there is no 
significant reduction in a margin of safety. The proposed changes 
will not alter any of the assumptions used in the accident analysis, 
nor will they cause any safety system parameters to exceed their 
acceptance limit. The proposed changes will not affect any 
operability requirements for equipment important to plant safety. 
Therefore, the proposed changes will not result in a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: November 23, 1999.
    Description of amendment request: The proposed changes will update 
the list of documents describing the analytical methods used to 
determine the core operating limits, specified in Technical 
Specification 6.9.1.8b.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change in document 1 of Technical Specification 
6.9.1.8b is made to provide the most recent, Nuclear Regulatory 
Commission (NRC) approved, methodology description and benchmarking 
results of the reactor analysis system used in the core neutronics 
analysis of cycle 14 and beyond. This change has no impact on plant 
equipment operation. Since the change only affects the neutronics 
analysis of the core, it cannot affect the likelihood or 
consequences of accidents. Therefore, this change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change in document 8 of Technical Specification 
6.9.1.8b is made to include the most recent, NRC approved, Emergency 
Core Cooling System (ECCS) model used in Large Break Loss of Coolant 
Accident (LBLOCA) applications. This model contains resolution of 
the deficiencies reported under 10 CFR 50.46(a) in a letter dated 
May 20, 1999. The use of the revised methodology also constitutes an 
improvement over the previous methodology. Therefore, this change 
will not significantly increase the probability or consequences of 
an accident previously evaluated.
    The proposed changes in document 4 of Technical Specification 
6.9.1.8b are administrative in nature. Therefore, these changes will 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change in document 1 of Technical Specification 
6.9.1.8b is made to provide the most recent, NRC approved, 
methodology description and benchmarking results of the reactor 
analysis system used in the neutronics analysis of cycle 14 and 
beyond. The proposed change in document 1 of Technical Specification 
6.9.1.8b will not alter the plant configuration (no new or different 
type of equipment will be installed) or require any new or unusual 
operator actions. It does not alter the way any structure, system, 
or component functions and does not alter the manner in which the 
plant is operated.
    The proposed change in the documents in number 8 of Technical 
Specification 6.9.1.8b is made to include the most recent, NRC 
approved, ECCS model used in LBLOCA applications. The proposed 
change in document 8 of Technical Specification 6.9.1.8b will not 
alter the plant configuration (no new or different type of equipment 
will be installed) or require any new or unusual operator actions. 
It does not alter the way any structure, system, or component 
functions and does not alter the manner in which the plant is 
operated.
    The proposed changes in document 4 of Technical Specification 
6.9.1.8b are administrative in nature. These changes do not alter 
the way any structure, system, or component functions and do not 
alter the manner in which the plant is operated.
    These changes do not introduce any new failure modes. Therefore, 
the proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change in document 1 of Technical Specification 
6.9.1.8b is made to provide the most recent, NRC approved, 
methodology description and benchmarking results of the reactor 
analysis system used in the neutronics analysis of cycle 14 and 
beyond. It has no impact on plant equipment operation. The proposed 
change in document 8 of Technical Specification 6.9.1.8b is made to 
include the most recent, NRC approved, ECCS model used in LBLOCA 
applications. This model contains resolution of the deficiencies 
reported under 10 CFR 50.46(a) in a letter dated May 20, 1999. The 
use of the revised methodology still provides a conservative 
simulation of the LBLOCA and conservative core neutronics analysis. 
The use of the revised methodology also constitutes an improvement 
over the previous methodology. The new documents will clearly 
identify the approved Siemens Topical Reports applicable to 
Millstone Unit No. 2 and will ensure that methodology changes will 
be identified and submitted to the NRC for approval, as required. 
The proposed changes in document 4 of Technical Specification 
6.9.1.8b are administrative in nature. Therefore, the proposed 
changes will not result in a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

[[Page 4285]]

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

    Date of amendment request: December 6, 1999.
    Description of amendment request: The proposed changes will modify 
the Technical Specification (TS) surveillance requirements associated 
with ensuring a limited number of charging and high pressure safety 
injection pumps are capable of injecting into the Reactor Coolant 
System when the plant is shutdown. In addition, the TS Bases will be 
modified to address these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed modifications to the surveillance requirements 
(SRs) associated with Technical Specifications 3.1.2.3, 3.1.2.4, and 
3.4.9.3 will remove information that specifies the methods to be 
used to perform the associated SRs. These SRs verify the maximum 
number of charging and high pressure safety injection (HPSI) pumps 
capable of injecting into the RCS [Reactor Coolant System] when the 
plant is shut down. This information will be transferred to the 
associated Bases. Additional methods associated with the charging 
pumps, which are technically equivalent to the current method, will 
be included in the Bases change. This will not change the 
requirement to verify that the associated pumps are not capable of 
injecting into the RCS when the plant is shut down.
    The proposed changes to the Technical Specifications and Bases 
will have no adverse effect on plant operation, or the availability 
or operation of any accident mitigation equipment. The plant 
response to the design basis accidents will not change. In addition, 
the proposed changes can not cause an accident. Therefore, there 
will be no significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Technical Specification and Bases changes will not 
alter the plant configuration (no new or different type of equipment 
will be installed) or require any new or unusual operator actions. 
They do not alter the way any structure, system, or component 
functions and do not significantly alter the manner in which the 
plant is operated. The proposed changes do not introduce any new 
failure modes. Also, the response of the plant and the operators 
following these accidents is unaffected by the changes. Therefore, 
the proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed modifications to the surveillance requirements 
associated with Technical Specifications 3.1.2.3, 3.1.2.4, and 
3.4.9.3 will remove information that specifies the methods to be 
used to perform the associated surveillance requirements. This will 
not change the requirement to verify that the associated pumps are 
not capable of injecting into the RCS when the plant is shut down.
    The proposed changes to the Technical Specifications and Bases 
will have no adverse effect on plant operation or equipment 
important to safety. The plant response to the design basis 
accidents will not change and the accident mitigation equipment will 
continue to function as assumed in the design basis accident 
analysis. Therefore, there will be no 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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: December 7, 1999.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) are associated with the action 
requirement to suspend positive reactivity additions. These changes 
will remove the action requirement to suspend positive reactivity 
additions from TS 3.4.2.1, ``Reactor Coolant System--Safety Valves,'' 
3.4.2.2, ``Reactor Coolant System--Safety Valves,'' and 3.7.6.1, 
``Plant Systems--Control Room Emergency Ventilation System,'' and 
provide guidance in the Bases for other TSs that require the suspension 
of positive reactivity addition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

Technical Specifications 3.4.2.1 and 3.4.2.2

    The proposed changes to Technical Specifications 3.4.2.1 and 
3.4.2.2, which address the pressurizer code safety valves in Modes 1 
through 4, will combine these two specifications into one Technical 
Specification, 3.4.2. The slight reduction in the Mode of 
Applicability for the new Technical Specification, to be consistent 
with the Mode of Applicability for Technical Specification 3.4.9.3, 
which addresses the Low Temperature Overpressure Protection (LTOP) 
System, is too small to result in a change in plant operations. The 
LCO [limiting condition for operation] for the pressurizer code 
safety valves in Mode 4 with all Reactor Coolant System (RCS) cold 
leg temperatures > 275  deg.F will be expanded to require all 
pressurizer code safety valves to be operable, instead of at least 
one pressurizer code safety valve. This more restrictive change will 
require additional accident mitigation equipment to be operable. The 
proposed action requirements for plant operation in Modes 1, 2, and 
3 have been expanded to require the plant to be in Mode 3 within 6 
hours and in Mode 4 within the following 6 hours, instead of just 
Mode 4 within 12 hours. In addition, the action requirements will be 
modified to address 2 inoperable pressurizer code safety valves. An 
entry into Technical Specification 3.0.3 will no longer be necessary 
if both pressurizer code safety valves are inoperable. In addition, 
the proposed action requirements are more restrictive than the 
action requirements of Technical Specification 3.0.3. The proposed 
action requirements for Mode 4 with all RCS cold leg temperatures > 
275  deg.F are different. The new Mode 4 action requirements will 
direct the plant to be cooled down to the applicability of Technical 
Specification 3.4.9.3, which will require the LTOP System to be 
placed in service to provide RCS overpressure protection. The 
proposed action requirements will ensure that the plant is placed in 
a condition where sufficient accident mitigation equipment will be 
available.
    The proposed Technical Specification, 3.4.2, will ensure the RCS 
has adequate overpressure protection when operating above 275 
deg.F. If the pressurizer code safety valves are not operable, the 
proposed Technical Specification will require a plant shutdown that 
will place the plant within the capability of the LTOP System to 
provide RCS overpressure protection. The proposed changes will have 
no adverse effect on plant operation, or the availability or 
operation of any accident mitigation equipment. The plant response 
to the design basis accidents will not change. In addition, the 
proposed changes can not cause an accident. Therefore, there will be 
no significant increase in the probability or consequences of an 
accident previously evaluated.

Technical Specification 3.7.6.1

    The proposed change to Technical Specification 3.7.6.1 will 
remove the requirement to suspend positive reactivity additions if 
both control room ventilation trains are inoperable in Modes 5 and 
6. The Control Room Ventilation System is required

[[Page 4286]]

to be operable in Modes 5 and 6 to protect the control room 
operators from an event that results in a rapid release of 
radioactivity, such as a fuel handling accident. In Modes 5 and 6, 
the positive reactivity addition methods of concern are boron 
dilution, RCS cooldown (negative isothermal temperature 
coefficient), and control rod withdrawal. Positive reactivity 
additions associated with fuel handling are already addressed by the 
additional action requirement in this specification to suspend core 
alterations. Control rod withdrawal is prohibited by Technical 
Specification 3.1.3.7, unless the RCS boron concentration is greater 
than or equal to the refueling boron concentration of Technical 
Specification 3.9.1. If the RCS is borated to the refueling 
concentration, sufficient negative reactivity has been added to 
compensate for the positive reactivity addition associated with 
control rod withdrawal in Modes 5 and 6. Therefore, only boron 
dilution and RCS temperature changes are of concern. However, both 
of these methods will result in slow changes to core reactivity in 
Modes 5 and 6, and since adequate shutdown margin (SDM) will have 
been established prior to entering Mode 5 or 6 (Technical 
Specifications 3.1.1.2 and 3.9.1), neither method will result in a 
rapid release of radioactivity. Therefore, the requirement to 
suspend positive reactivity additions is not necessary for the 
protection of the control room operators.
    The proposed change will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. In addition, the proposed change can not 
cause an accident. Therefore, there will be no significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Technical Specification will not alter the plant 
configuration (no new or different type of equipment will be 
installed) or require any new or unusual operator actions. They do 
not alter the way any structure, system, or component functions and 
do not significantly alter the manner in which the plant is 
operated. The proposed changes do not introduce any new failure 
modes. Also, the response of the plant and the operators following 
these accidents is unaffected by the changes. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to combine Technical Specifications 3.4.2.1 
and 3.4.2.2 into a new Technical Specification, 3.4.2, will result 
in a slight reduction in the Mode of Applicability for the new 
Technical Specification, will require both pressurizer code safety 
valves to be operable in Mode 4 with all RCS cold leg temperatures > 
275  deg.F, will modify the action requirements in Modes 1, 2, and 3 
to add a requirement to be in Mode 3 within 6 hours and to address 
two inoperable pressurizer code safety valves, and will provide 
different action requirements for Mode 4 with all RCS cold leg 
temperatures > 275  deg.F. The reduction in Mode of Applicability is 
too small to adversely impact plant operations. Requiring both 
pressurizer code safety valves to be operable in Mode 4 with all RCS 
cold leg temperatures > 275  deg.F will provide additional accident 
mitigation equipment. The modified action requirement to be in Mode 
3 within 6 hours will not change the requirement to be in Mode 4 
within 12 hours. The action requirements added to address two 
inoperable pressurizer code safety valves are more restrictive than 
the action requirements of Technical Specification 3.0.3. The new 
Mode 4 action requirements will direct the plant to be cooled down 
to the applicability of Technical Specification 3.4.9.3, which will 
require the LTOP System to be placed in service to provide RCS 
overpressure protection. The proposed action requirements will 
ensure that the plant is placed in a condition where sufficient 
accident mitigation equipment will be available.
    The proposed change to Technical Specification 3.7.6.1 will 
remove the requirement to suspend positive reactivity additions if 
both control room ventilation trains are inoperable in Modes 5 and 
6. The Control Room Ventilation System is required to be operable in 
Modes 5 and 6 to protect the control room operators from an event 
that results in a rapid release of radioactivity, such as a fuel 
handling accident. The proposed change will only impact slow methods 
to change core reactivity, such as boron dilution and RCS 
temperature changes. Therefore, the action requirement to suspend 
positive reactivity additions is not necessary for the protection of 
the control room operators.
    The proposed changes will have no adverse effect on plant 
operation or equipment important to safety. The plant response to 
the design basis accidents will not change and the accident 
mitigation equipment will continue to function as assumed in the 
design basis accident analysis. Therefore, there will be no 
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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of amendment request: November 23, 1999.
    Description of amendment request: The proposed change affects 
Technical Specification 4.0.5, ``Limiting Conditions for Operation and 
Surveillance Requirements'' by adding a biennial or 2-year surveillance 
interval and incorporating a required frequency for performing 
inservice testing activities of once per 731 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change amends Technical Specification Section 
4.0.5.b by adding a biennial or 2 year surveillance to the existing 
list. This surveillance interval is included as part of the current 
Millstone Unit Nos. 2 and 3 Inservice Test (IST) surveillance 
program. Inclusion of this surveillance interval in the facility 
Technical Specifications clarifies the applicability of this 
surveillance interval and affords operational flexibility in the 
event a surveillance cannot be completed within the required 
interval.
    The proposed change will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. In addition, the proposed change can not 
cause an accident. Therefore, there will be no significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The biennial surveillance relates to performing inservice 
testing of plant components. The possibility of a new or different 
kind of accident from any accident previously evaluated is not 
created because the proposed Technical Specification change does not 
introduce a new mode of plant operations and does not involve 
physical modifications to the plant. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    There is no impact on the margin of safety as defined in the 
Technical Specifications. Performance of surveillance tests at 
regular intervals provides assurance of reliability and availability 
of accident mitigating equipment. The Technical Specifications 
provide the required frequency for performing surveillance testing. 
Adding a new surveillance frequency to the Technical Specifications 
will provide consistent yet acceptable flexibility in scheduling 
surveillance tests and provide additional assurance that testing 
will be performed in a timely manner.
    The proposed change will have no adverse effect on plant 
operation or equipment

[[Page 4287]]

important to safety. The plant response to the design basis 
accidents will not change and the accident mitigation equipment will 
continue to function as assumed in the design basis accident 
analysis. Therefore, there will be no 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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: November 29, 1999.
    Description of amendment request: The requested changes would 
revise Technical Specification (TS) 3/4.6.6, ``Supplementary Leak 
Collection and Release System,'' (SLCRS), TS 3/4.7.7, ``Control Room 
Emergency Ventilation System,'' (CREVS), TS 3/4.7.9, ``Auxiliary 
Building Filter System,'' (ABFS), and 3/4.9.12, ``Fuel Building Exhaust 
System,'' (FBES), in response to Generic Letter (GL) 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The 
requested changes require testing of nuclear-grade activated charcoal 
to be conducted in accordance with American Society for Testing 
Materials (ASTM) D3803-1989, ``Standard Test Method for Nuclear-Grade 
Activated Carbon,'' as recommended by GL 99-02.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    NNECO [Northeast Nuclear Energy Company] has reviewed the 
proposed revision in accordance with 10 CFR 50.92 and has concluded 
that the revision does not involve any Significant Hazards 
Consideration (SHC). The basis for this conclusion is that the three 
criteria of 10 CFR 50.92(c) are not satisfied. The proposed TS 
revision does not involve an SHC because the revision would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change modifies the TS to reference ASTM D3803-
[19]89 for performing laboratory testing of nuclear-grade charcoal 
in ESF [Engineered Safeguards Features] filtration systems. The 
testing methodology associated with ASTM D3803-[19]89 provides more 
stringent requirements than what is currently employed. These more 
stringent requirements, along with a factor of safety of greater 
than or equal to two in regards to the charcoal efficiency assumed 
in the design bases dose analysis will not result in operations that 
will increase the probability of initiating an analyzed event and do 
not alter assumptions relative to mitigation of an accident or 
transient event. The more restrictive requirements continue to 
ensure process variables, structures, systems, and components are 
maintained consistent with the safety analyses and licensing basis. 
There are no related modifications to any systems. The proposed 
change does not affect procedures governing plant operations. 
Therefore there is no significant increase in the probability [or 
consequences] of occurrence of a previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change modifies the TS to reference ASTM D3803-
[19]89 for performing laboratory testing of nuclear-grade charcoal 
in ESF filtration systems. The proposed change does not involve the 
physical alteration of the plant (no new or different type of 
equipment will be installed) or changes in the methods governing 
normal plant operation. This change does impose different, more 
conservative testing requirements on the ESF filtration system 
charcoal samples. However there is no alteration in the methods 
employed to obtain the charcoal sample and testing is performed 
offsite. These changes are consistent with the safety analyses and 
licensing basis. Furthermore, the proposed changes do not introduce 
any new modes of plant operation, or alter any operational 
setpoints. Thus the possibility of a new or different kind of 
accident from any previously evaluated is not created.
    3. Involve a significant reduction in the margin of safety.
    The proposed change modifies the TS to reference ASTM D3803-
[19]89 for performing laboratory testing of nuclear-grade charcoal 
in ESF filtration systems. The imposition of the more conservative 
charcoal filter testing requirements associated with ASTM D3803-
[19]89 along with a factor of safety of greater than or equal to 
two, in regards to the charcoal efficiency assumed in the design 
bases dose analysis has no impact on, nor decreases the margin of 
plant safety. The conservative nature of ASTM D3803-[19]89 is by 
definition, providing additional restrictions to enhance plant 
safety. This change maintains requirements within the safety 
analysis and licensing basis. Therefore, there will be no 
significant reduction in the margin of safety as defined in the 
Bases for the TS affected by the proposed change.
    As described above this TSCR [Technical Specification Change 
Request] does not impact the probability of an accident previously 
evaluated, does not involve a significant increase in the 
consequences of an accident previously evaluated, does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, and does not result in a significant 
reduction in a margin of safety. Therefore, NNECO has concluded that 
the proposed changes do not involve an SHC.

    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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: November 5, 1999.
    Description of amendment request: The proposed changes will revise 
LGS Technical Specifications (TSs) to incorporate revised testing and 
acceptance criteria for the performance of laboratory analysis of 
safety-related nuclear-grade activated charcoal in response to Generic 
Letter (GL) 99-02. ``Laboratory Testing of Nuclear-Grade Activated 
Charcoal,'' dated June 3, 1999. In addition, minor editorial changes 
are being proposed for wording consistency and to correct a 
typographical error.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Changing the methodology for the performance of the laboratory 
testing of nuclear-grade activated charcoal samples from Reg. 
[Regulatory] Guide 1.52 to ASTM D3803-1989 in accordance with 
Generic Letter 99-02, and establishing a new methyl iodide 
penetration acceptance criteria does not involve any physical 
changes or modifications to the function or operation of any safety-
related structure, system, or component. The new testing methodology 
will enable a more accurate, conservative and reliable determination 
of the charcoal decontamination efficiencies associated with the 
SGTS [Standby Gas Treatment System], RERS [Reactor Enclosure 
Recirculation System], and CREFAS [Control Room Emergency Fresh Air 
System] which will better assure that the assumed charcoal 
efficiencies credited in the licensed accident

[[Page 4288]]

analysis are adequately maintained. Implementing this change will 
only involve revisions to existing procedures.
    The SGTS, RERS, and CREFAS are standby systems that are designed 
to mitigate the consequences of the analyzed accidents. No analyzed 
accident initiating events are impacted, no new accident initiators 
or new failure modes are created and the credited charcoal 
efficiency for each system in the licensed accident analyses is not 
changing as a result of the proposed changes. The ability of the 
SGTS, RERS, and CREFAS to perform all of their safety-related 
mitigation functions as designed will not be affected by the 
proposed changes. Furthermore, the change in the testing methodology 
and acceptance criteria will not result in increasing the dose rates 
currently calculated in the existing accident analyses.
    In addition, the proposed minor editorial changes are 
administrative in nature and do not impact the operation, physical 
configuration, or function of plant equipment or systems. The 
proposed editorial changes do not impact the initiators or 
assumptions of analyzed events, nor do they impact mitigation of 
accidents or transient events.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Changing the methodology for the performance of the laboratory 
testing of nuclear-grade activated charcoal in accordance with 
Generic Letter 99-02, and establishing new methyl iodide penetration 
acceptance criteria is not an accident initiator, does not create 
any new failure modes, nor does it result in the occurrence of an 
accident. This change does not result in any physical plant 
modification and does not affect the safety-related function, 
assigned charcoal efficiency assumed in the accident analyses, or 
operation of the SGTS, RERS, and CREFAS. This change will only 
involve revisions to existing procedures.
    In addition, the proposed minor editorial changes are 
administrative in nature and do not alter plant configuration, 
require that new equipment be installed, alter assumptions made 
about accidents previously evaluated, or impact the operation or 
function of plant equipment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The safety-related air cleaning units used in ESF [Engineered 
Safety Feature] ventilation systems reduce the potential onsite and 
offsite consequences of a radiological accident by adsorbing 
radioiodine. Changing the methodology for the performance of the 
laboratory testing of nuclear-grade activated charcoal samples from 
Reg. Guide 1.52 to ASTM D3803-1989 in accordance with Generic Letter 
99-02, and the establishment of new methyl iodide penetration 
acceptance criteria does not increase the dose rates above what is 
currently calculated in the accident analyses.
    In addition, the proposed minor editorial changes are 
administrative in nature and do not involve any physical changes to 
plant structures, systems or components (SCCs), or the manner in 
which SSCs are operated, maintained, modified, tested, or inspected. 
The proposed editorial changes do not involve a change to any safety 
limits, limiting safety system settings, limiting conditions of 
operation, or design parameters for any SSC. The proposed editorial 
changes do not impact any safety analysis assumptions and do not 
involve a change in initial conditions, system response times, or 
other parameters affecting any accident analysis.
    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: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: November 17, 1999.
    Description of amendment request: The proposed changes will revise 
the Peach Bottom Units 2 and 3 Technical Specifications (TSs) Section 
5.5.7.c., Ventilation Filter Testing Program (VFTP), in accordance with 
Generic Letter (GL) 99-02, ``Laboratory Testing of Nuclear-Grade 
Activated Charcoal.'' This TS change will (1) specify that the 
laboratory testing for methyl iodide penetration be performed 
referencing ASTM D3803-1989 at a temperature of 30  deg.C (86  deg.F), 
and (2) revise the acceptance criteria for methyl iodide penetration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Changing the methodology for the performance of the laboratory 
testing of nuclear grade activated charcoal samples from RG 
[Regulatory Guide] 1.52 to ASTM [American Society for Testing and 
Materials] D3803-1989 and the establishment of new methyl iodide 
penetration acceptance criteria and test temperature in accordance 
with Generic Letter 99-02, do not involve any changes or 
modifications to the function or operation of any safety related 
structure, system, or component. The new testing methodology enables 
a more accurate and conservative charcoal decontamination efficiency 
to be determined which better assures that the assumed charcoal 
efficiency credited in the licensed accident analysis is being 
adequately maintained. Implementing this change only involves 
revisions to existing procedures.
    The SGTS [Standby Gas Treatment System] and MCREVS [Main Control 
Room Emergency Ventilation System] are standby systems that are 
designed to mitigate the consequences of the analyzed accidents. No 
analyzed accident initiating events are impacted, no new accident 
initiators or new failure modes are created and the credited 
charcoal efficiency for each system in the licensed accident 
analyses is not changing. The change in laboratory testing 
methodology does not degrade the ability of these systems to perform 
all of their safety related mitigation functions as designed.
    Therefore, the proposed changes described above do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Changing the methodology for the performance of the laboratory 
testing of nuclear grade activated charcoal in accordance with 
Generic Letter 99-02 and establishing new methyl iodide penetration 
acceptance criteria is not an accident initiator, does not create 
any new failure modes, nor does it result in the occurrence of an 
accident. This change does not result in any physical plant 
modification and does not affect the safety related function, 
charcoal efficiency, or operation of the SGTS or MCREVS. This change 
only involves revisions to existing procedures to comply with NRC 
guidance from GL 99-02.
    Therefore, the possibility of a new or different kind of 
accident than previously evaluated is not created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The safety related air cleaning units used in ESF [Engineered 
Safety Feature] ventilation systems reduce the potential onsite and 
offsite consequences of a radiological accident by absorbing 
radioiodine. Changing the methodology for the performance of the 
laboratory testing of nuclear-grade activated charcoal samples from 
RG 1.52 to ASTM D3803-1989 in accordance with Generic Letter 99-02, 
and the establishment of new methyl iodide penetration acceptance 
criteria does not increase the dose rates above what is currently 
calculated in the accident analyses.
    Therefore, the above change does not involve a significant 
reduction in a margin of safety.


[[Page 4289]]


    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: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: November 16, 1999.
    Description of amendment request: The proposed amendment would 
revise the Trojan Nuclear Plant (TNP) Permanently Defueled Technical 
Specifications by removing Figure 4.1-1, ``Site and Exclusion Area 
Boundaries,'' from Section 4.0, ``Design Features,'' and incorporate 
the applicable portion of this figure in the Trojan Nuclear Plant 
Defueled Safety Analysis Report. Other associated administrative 
changes resulting from the deletion of Figure 4.1-1, as well as an 
editorial change to the table of contents, 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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The requested license amendment consists of changes that are 
administrative and/or editorial in nature, in that the physical and 
operational characteristics of the TNP site are unchanged. As such, 
the requested amendment does not in any way affect systems, 
structures, or components that could initiate or be required to 
mitigate the consequences of an accident previously evaluated. 
Therefore, the proposed 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 requested license amendment consists of changes that are 
administrative and/or editorial in nature, in that the physical and 
operational characteristics of the TNP site are unchanged. As such, 
the requested amendment does not affect systems, structures, or 
components in any way not previously evaluated, and no new or 
different failure modes will be created. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The requested license amendment consists of changes that are 
administrative and/or editorial in nature, in that the physical and 
operational characteristics of the TNP site are unchanged. 
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: Douglas R. Nichols, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Section Chief: Michael T. Masnik.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of amendment request: December 27, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) 4.6.2.2.b, ``Suppression Pool 
Spray,'' and 4.6.2.3.b, ``Suppression Pool Cooling,'' to modify the 
acceptance criteria associated with flow rate testing of the Residual 
Heat Removal (RHR) system pumps.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS change does not involve any physical changes to 
plant structures, systems or components (SSC). The RHR system will 
continue to function as designed. The RHR system is designed to 
mitigate the consequences of an accident, and therefore, cannot 
contribute to the initiation of any accident. The proposed TS 
surveillance requirement changes implement testing methods that more 
appropriately control and reflect RHR operation and establish 
acceptance criteria, which ensure that Hope Creek's licensing and 
design basis assumptions are met. In addition, this proposed TS 
change will not increase the probability of occurrence of a 
malfunction of any plant equipment important to safety, since the 
manner in which the RHR system is operated is not affected by these 
proposed changes. The proposed surveillance requirement acceptance 
criteria ensure that the RHR safety functions will be accomplished. 
Therefore, the proposed TS changes would not result in the increase 
of the consequences of an accident previously evaluated, nor do they 
involve an increase in the probability 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 TS changes do not involve any physical changes to 
the design of any plant SSC. The design and operation of the RHR 
system is not changed from that currently described in Hope Creek's 
licensing basis. The RHR system will continue to function as 
designed to mitigate the consequences of an accident. Implementing 
the proposed changes does not result in plant operation in a 
configuration that would create a different type of malfunction to 
the RHR system than any previously evaluated. In addition, the 
proposed TS changes do not alter the conclusions described in Hope 
Creek's licensing basis regarding the safety related functions of 
this system.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes contained in this submittal would implement 
testing methods that adequately demonstrate RHR pump capability and 
establish acceptance criteria consistent with Hope Creek's licensing 
basis. The ability of RHR to perform its safety functions is not 
adversely affected by these proposed changes. Therefore, the 
proposed TS 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: December 29, 1999.
    Description of amendment request: The proposed amendments would 
revise the Salem Nuclear Generating Station Technical Specification 
requirements for instrumentation in the reactor trip system by adding 
tolerances to certain setpoint values.

[[Page 4290]]

    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The accidents of concern affected by the over-temperature or 
over-power delta temperature [trip signal] which have been evaluated 
are unaffected by the proposed editorial changes thus the changes do 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The changes proposed are editorial in nature and do not alter 
physical configuration, replace or modify existing equipment, affect 
operating practices or create any new or different accident 
precursors which could impact on the accident analysis. Thus there 
is no possibility of a new or different kind of accident as a result 
of the proposed changes.
    3. Does not involve a significant reduction in a margin of 
safety.
    No margin of safety will be reduced by the proposed changes. The 
proposed changes do not adversely affect the ability of the trip 
systems to operate when called upon. Rather, these changes should 
result in clarity regarding the proper calibration of the trip 
instrumentation and therefore the margin of safety is preserved for 
those events in which there is a dependence upon an over-temperature 
or over-power delta temperature trip signal.

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

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: November 30, 1999.
    Description of amendment request: The proposed amendment would 
allow the Rochester Gas and Electric Corporation to revise Sections 
5.5.10 (a.3), (c.5), and (d.3) of the Ginna Station Improved Technical 
Specifications (ITS) to provide a reference to American Society for 
Testing and Materials (ASTM) Standard Procedure D3803-1989 as the 
procedure for performing laboratory testing of charcoal adsorbers that 
are installed in the Ginna Control Room Emergency Air Treatment System 
(CREATS), Containment Post-Accident Sampling System (CPASS), and Spent 
Fuel Pool Charcoal Absorber System (SFPCAS). These charcoal adsorbers 
for the CREATS and CPASS are installed for the purpose of reducing the 
levels of radioactive iodide species released to the containment and 
control room during a postulated design basis, while the charcoal 
adsorbers in the SFPCAS are installed for reducing the levels of 
radioactive iodide species released to the auxiliary building during a 
postulated fuel handling accident. The changes to ITS Sections (a.3), 
(c.5), and (d.3) will also provide a specific test temperature and 
humidity level for performing the testing of the charcoal adsorbers, 
and to increase the allowable penetration of methyl iodide to these 
systems from 10% to 14.5%. The requests for the changes are consistent 
with the staff's position stated in NRC Generic Letter 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated June 
3, 1999.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. With respect to the more restrictive proposals 
associated with providing a reference to ASTM D3803-1989, ``Standard 
Test Method for Nuclear-Grade Activated Carbon,'' and providing a 
specific test temperature and relative humidity for testing the 
charcoal adsorbers, the proposed changes do not involve a significant 
hazards consideration as discussed below:

    (1) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The changes add 
a reference to the latest approved test protocol and provide for 
specific test conditions. This does not increase the probability of 
an accident previously evaluated since the tests are of themselves 
not an accident initiator. The proposed changes are in accordance 
with NUREG-1431 guidance and provide a higher assurance of the 
ability of the charcoal adsorbers to perform as assumed in the 
accident analysis. Therefore, the probability or consequences of an 
accident previously evaluated is not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes add specific details of charcoal adsorber testing and do not 
of themselves involve a physical alteration of the plant (ie. no new 
or different type of equipment will be added to perform the required 
testing) or changes in the methods governing normal plant operation. 
The changes only involve implementing currently approved test 
methodology. Therefore, the possibility for a new or different kind 
of accident from any accident previously evaluated is not created.
    (3) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes only add conservatism in the test 
requirements for the charcoal adsorbers credited in the accident 
analysis. ASTM D3803-1989 is considered to be the most accurate and 
most realistic protocol for testing charcoal in ventilation systems 
because it offers the greatest assurance of accurately and 
consistently determining the capability of the charcoal. Therefore, 
this change does not involve a significant reduction in a margin of 
safety.
    With respect to the less restrictive proposal to increase the 
allowable test limit for methyl iodide penetration of charcoal 
adsorbers, the changes do not involve a significant hazards 
consideration as discussed below:
    (4) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The changes 
revise the acceptance criteria for the allowed penetration of methyl 
iodide during the testing of charcoal adsorbers in the plant 
ventilation systems. This does not increase the probability of an 
accident previously evaluated since the tests are of themselves not 
an accident initiator. Because ASTM D3803-1989 is a more accurate 
and demanding test than older tests this new protocol will allow the 
use a safety factor of 2 for determining the acceptance criteria for 
charcoal filter efficiency. The new acceptance criteria continue to 
ensure that the efficiency assumed in the accident analysis is still 
valid. Therefore, the probability or consequences of an accident 
previously evaluated is not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes of revising charcoal adsorber testing acceptance criteria do 
not of themselves involve a physical alteration of the plant (ie. no 
new or different type of equipment will be added to perform the 
required testing) or changes in the methods governing normal plant 
operation. Therefore, the possibility for a new or different kind of 
accident from any accident previously evaluated is not created.
    (3) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes only revise the test acceptance 
criteria of charcoal adsorbers as the result of implementing testing 
in accordance with ASTM D3803-1989. ASTM D3803-1989 is considered to 
be the most accurate and most realistic protocol for testing 
charcoal in ventilation systems because it offers the greatest 
assurance of

[[Page 4291]]

accurately and consistently determining the capability of the 
charcoal. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    Based upon the preceding information, the Rochester Gas and 
Electric Corporation determined that the proposed changes do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated, create the possibility of a new or 
different kind of accident from any accident previously evaluated, or 
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: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Section Chief: Marsha Gamberoni, Acting.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: September 30, 1999 (TS 98-005).
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Plant Unit 1 Technical Specifications (TS) 
analytical methods for core operating limits to implement an analysis 
supporting a more negative moderator temperature coefficient (MTC) for 
the end of cycle condition. This alternate methodology is based on a 
Westinghouse Electric Company analysis documented in reports WCAP-
15088-P, Revision 1 (proprietary), ``Safety Evaluation Supporting a 
More Negative EOL Moderator Temperature Coefficient Technical 
Specification for the Watts Bar Nuclear plant,'' and WCAP-15099-P, 
Revision 1 (non-proprietary, same title).
    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 more negative EOL [end-of-life] MTC does not increase the 
probability of an accident previously evaluated in the FSAR [Final 
Safety Analysis Report]. No new performance requirements are being 
imposed on any system or component such that any design criteria 
will be exceeded. The conservative MDC [moderator density 
coefficient] assumption in the current analyses of record has been 
confirmed to remain bounding for the more negative proposed TS 
values. Therefore, no change in the modeling of the accident 
analysis conditions or response is necessary in order to implement 
this change. The consequences of an accident previously evaluated in 
the FSAR are not increased due to the more negative EOL MTC. The 
dose predictions presented in the FSAR remain valid such that no 
more severe consequences will result.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The more negative EOL MTC does not create the possibility of an 
accident which is different than any already evaluated in the FSAR. 
No new failure modes have been defined for any system or component 
nor has any new limiting single failure been identified. 
Conservative assumptions for MDC have already been modeled in the 
FSAR analyses and it has been determined that the more negative MTC 
values to be implemented in the TS will continue to be bounded by 
these assumptions.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The evaluation of the more negative EOL MTC has taken into 
account the applicable technical specifications and has bounded the 
conditions under which the specifications permit operation. The 
applicable technical specification is Section 5.9.5.b which lists 
methods approved by the NRC for use in determining the core 
operating limits. The values of the LCO [limiting condition for 
operation] and SRs [surveillance requirements] are located in the 
COLR [core operating limits report]. The analyses which support 
these technical specifications have been evaluated. The results as 
presented in the FSAR remain bounding for the more negative EOL MTC. 
Therefore, the margin of safety, as defined in the bases to these 
technical specifications, 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard Correia.

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

    Date of amendment request: December 21, 1999.
    Description of amendment request: This proposed change revises the 
control rod block requirements consistent with the BWR/4 Standard 
Technical Specifications. Some functions are proposed to be relocated 
to the Technical Requirements Manual, the requirements for the retained 
functions are clarified, and two functions are added to the Technical 
Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The relocated functions are not assumed as initial conditions 
for, nor are they credited in the mitigation of, any design basis 
accident or transient previously evaluated. Since reactor operation 
with these revised and relocated Specifications is fundamentally 
unchanged, no design or analytical acceptance criteria will be 
exceeded. As such, this change does not impact initiators of 
analyzed events nor assumed mitigation of design basis accident or 
transient events.
    More stringent and purely administrative changes do not affect 
the initiation of any event, nor do they negatively impact the 
mitigation of any event. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    None of the proposed changes affects any parameters or 
conditions that could contribute to the initiation of an accident. 
No new accident modes are created since the manner in which the 
plant is operated is unchanged. No safety-related equipment or 
safety functions are altered as a result of these changes. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    There is no impact on equipment design or operation, and there 
are no changes being made to safety limits or safety system settings 
that would adversely affect plant safety as a result of the proposed 
changes. Since the changes have no effect on any safety analysis 
assumption or initial condition, the margins of safety in the safety 
analyses are maintained. In addition, neither administrative changes 
with no technical impact, nor the imposition of more stringent 
requirements have a negative impact on a margin of safety. 
Therefore, the proposed

[[Page 4292]]

changes do not involve a significant reduction in a margin of 
safety.

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

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

    Date of amendment request: December 15, 1999 (ET 99-0050).
    Description of amendment request: The proposed amendment would 
modify the Improved Technical Specifications (ITSs) that were issued in 
Amendment No. 123 on March 31, 1999, and implemented on December 18, 
1999. The proposed changes would expand the region of acceptable seal 
injection flow in Figure 3.5.5-1 of ITS 3.5.5 and provide the following 
10 editorial changes: (1) delete the redundant ``%'' sign in the 
allowable value for function 4 in Table 3.3.1-1 on reactor trip system 
instrumentation, (2) delete the extra spacing in the description of 
function 20 in Table 3.3.1-1, (3) insert periods at the end of the text 
for Conditions M and N in the actions for limiting condition for 
operation (LCO) 3.3.2 on engineered safety features actuation system 
instrumentation (ESFASI), (4) spell ``requirements'' correctly in 
function 5.c of Table 3.3.2-1 for ESFASI, (5) delete the unneeded ``SR 
3.3.2.6'' from the surveillance requirements column for Function 7.a in 
Table 3.3.2-1, (6) align the wording ``Coincident with Safety 
Injection'' with the title of Function 7.b in Table 3.3.2-1, (7) align 
the data in the 4 columns of Table 3.3.7-1, CREVS [control room 
emergency ventilation system] Actuation Instrumentation, for Function 3 
with the first line of the title of the function, (8) align the 
specified completion time in Condition B of the actions for LCO 3.7.1 
for main steam safety valves with text for the Required Action B.2, (9) 
add the acronym ``EES'' to Emergency Exhaust System in the table of 
contents and use the acronym in the upper right-hand-corner of the 4 
ITS pages for LCO 3.7.13 on the emergency exhaust system, and (10) 
uncapitalize the word ``Associated'' in Condition B of the actions for 
LCO 3.8.4 on DC sources--operating because it should not be 
capitalized. The licensee would also add text to the Bases to the 
applicable safety analyses for the seal injection flow of LCO 3.5.5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The restriction on RCP [reactor coolant pump] seal injection 
flow limits the amount of ECCS [emergency core cooling system] flow 
that would be diverted from the injection path following an 
accident. This limit is based on safety analysis assumptions that 
are required because RCP seal injection flow is not isolated during 
SI [safety injection]. The intent of the LCO 3.5.5 limit on seal 
injection flow is to make sure that flow through the RCP seal water 
injection line is low enough to ensure that sufficient centrifugal 
charging pump injection flow is directed to the RCS [reactor coolant 
system] via the injection points. The expansion of the Acceptable 
Range for the flow limits does not impact the assumed ECCS flow that 
would be available for injection into the RCS following an accident.
    There are no hardware changes nor are there any changes in the 
method by which any safety related plant system performs its safety 
function. Since the change continues to ensure 100 percent of the 
assumed charging flow is available, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed editorial changes involve corrections to the 
improved Technical Specifications that are associated with the 
original conversion application and supplements or the certified 
copy of the improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements of the Technical 
Specifications.
    Therefore, the proposed 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 of 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. 
The expansion of the Acceptable Range for the [seal injection] flow 
limits does not impact the assumed ECCS flow that would be available 
for injection into the RCS following an accident.
    The proposed editorial changes involve corrections to the 
improved Technical Specifications that are associated with the 
original conversion application and supplements or the certified 
copy of the improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements of the Technical 
Specifications.
    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 change [for seal injection flow] does not affect 
the acceptance criteria for any analyzed event. There will be no 
effect on the manner in which safety limits or limiting safety 
system settings are determined nor will there be any effect on those 
plant systems necessary to assure the accomplishment of protection 
functions. The expansion of the Acceptable Range for the flow limits 
does not impact the assumed ECCS flow that would be available for 
injection into the RCS following an accident.
    The proposed editorial changes involve corrections to the 
improved Technical Specifications that are associated with the 
original conversion application and supplements or the certified 
copy of the improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements of the Technical 
Specifications.
    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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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 4293]]

page cited. This notice does not extend the notice period of the 
original notice.

Indiana Michigan Power Company, Docket, Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: December 22, 1999.
    Brief description of amendments: The amendments would delete the 
Donald C. Cook (D.C. Cook), Unit 1 and 2, Technical Specification (TS) 
5.4.2, ``Reactor Coolant System Volume,'' because the information 
regarding the reactor coolant system (RCS) is not required by TS 
Section 5.0, ``Design Features,'' for compliance with 10 CFR 
50.36(c)(4). Changes to the RCS volume information are included in the 
D.C. Cook Updated Final Safety Analyses Report, and are controlled in 
accordance with 10 CFR 50.59.
    Date of publication of individual notice in Federal Register: 
January 13, 1999 (65 FR 2199).
    Expiration date of individual notice: February 14, 2000.

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

    Date of amendment request: March 31, 1999, as supplemented by 
letters dated May 20, June 1, July 14, and October 14, 1999.
    Description of amendment request: The amendment converts the 
current Technical Specifications (TSs) for the James A. FitzPatrick 
Nuclear Power Plant, to a set of improved TSs based upon NUREG-1433, 
``Standard Technical Specifications for General Electric Plants BWR/4'' 
Revision 1 dated April 1995.
    Date of publication of individual notice in Federal Register: 
November 8, 1999 (64 FR 60854).
    Expiration date of individual notice: December 8, 1999.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: June 11, 1999.
    Brief description of amendment: The amendment made various title 
changes to the plant organization.
    Date of issuance: January 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 219.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38027).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 7, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of application for amendments: October 2, 1998, as 
supplemented by letters dated April 13, 1999, and September 15, 1999. 
Information in Commonwealth Edison correspondence dated July 8, 1999, 
and August 30, 1999, was also considered during the review of the 
amendments.
    Brief description of amendments: The amendments replace the custom 
operational technical specifications with a set of permanently defueled 
technical specifications that reflect the permanently shutdown and 
defueled status of the Zion Nuclear Power Station, Units 1 and 2. The 
amendments also delete certain license conditions from the operating 
licenses that are no longer applicable to the facility in its 
permanently shutdown and defueled condition. Information supplied in 
Commonwealth Edison letters dated July 8, 1999, August 30, 1999, and 
September 15, 1999, provided clarifying information and did not expand 
the scope of the original Federal Register notice dated June 2, 1999, 
and did not change the staff's proposed no significant hazards finding.
    Date of issuance: December 30, 1999.
    Effective date: December 30, 1999.
    Amendment Nos.: Unit 1--180; Unit 2--167.
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications and the operating licenses.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29709).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 30, 1999.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: September 10, 1999 (NRC-99-
0072), as supplemented November 19, 1999 (NRC-99-0107).
    Brief description of amendment: The amendment revises the Technical 
Specification surveillance requirements for the Division I 130/260-volt 
dc battery to accommodate the design of the replacement battery.
    Date of issuance: January 12, 2000.
    Effective date: As of the date of issuance and shall be implemented 
prior to the startup from the seventh refueling outage.
    Amendment No.: 136.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59800). The November 19, 1999, letter provided clarifying information 
that was within

[[Page 4294]]

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 January 12, 2000.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: September 16, 1999, 
supplemented November 3, 1999.
    Brief description of amendments: The amendments revise Section 
3.8.4, ``DC Sources--Operating'' of the Technical Specifications. 
Specifically, the amendments modify Surveillance Requirements (SRs) 
3.8.4.8 and 3.8.4.9 and the associated Bases SR 3.8.4.8 and 3.8.4.9 to 
allow testing of the direct current (dc) channel batteries with the 
units on line. The change to SR 3.8.4.8 would also prohibit the diesel 
generator batteries from being service tested while the units are on 
line.
    Date of issuance: January 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment Nos.: Unit 1-183; Unit 2-175.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56529). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 7, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 9, 1999.
    Brief description of amendment: This amendment revised the Perry 
Nuclear Power Plant Environmental Protection Plan by eliminating the 
requirement to sample Lake Erie sediment in the Perry and Eastlake 
Plant area for Corbicula, since Corbicula and zebra mussels have 
already been identified, and control and treatment plans have been 
implemented which are effective for both species.
    Date of issuance: January 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 110.
    Facility Operating License No. NPF-58: This amendment revised the 
Environmental Protection Plan.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59802). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 5, 2000.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: November 3, 1999.
    Brief description of amendments: The amendments allow use of fuel 
rods with ZIRLO cladding, specify an alternate methodology to determine 
the integral fuel burnable absorber (IFBA) requirements for 
Westinghouse fuel assemblies stored in the new fuel storage racks, and 
delete the designation of the fuel assembly types allowed in the spent 
fuel storage racks and the new fuel storage racks.
    Date of issuance: January 6, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 239 and 220.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67335).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 6, 2000.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: March 18, 1998, as supplemented 
by letters dated March 25, September 29, and November 3, 1999.
    Brief description of amendments: The amendments change the way 
passive failures in the auxiliary saltwater (ASW) and component cooling 
water (CCW) systems are mitigated during the long-term recovery period 
following a loss-of-coolant accident (LOCA). Specifically, plant 
procedures will no longer require ASW and CCW system train separation 
after the transfer to hot leg recirculation following a LOCA.
    Date of issuance: January 13, 2000.
    Effective date: January 13, 2000, and shall be implemented in the 
next periodic update to the FSAR Update in accordance with 10 CFR 
50.71(e).
    Amendment Nos.: Unit 1--138, Unit 2--138.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Final Safety Analysis Report Update.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53953).
    The supplemental letters dated March 25, September 29, and November 
3, 1999, provided additional clarifying information, did not expand the 
scope of the application as originally noticed, and did not change the 
staff's initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 13, 2000.
    No significant hazards consideration comments received: No.

PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit 
1, Luzerne County, Pennsylvania

    Date of application for amendment: March 12, 1999, as supplemented 
by letter dated November 1, 1999.
    Brief description of amendment: This amendment revised the Minimum 
Critical Power Ratio safety limits in TS Section 2.1.1.2 and modified 
the references in TS Section 5.6.5 of a critical power correlation 
applicable to Siemens Power Corporation Atrium-10 fuel.
    Date of issuance: December 30, 1999.
    Effective date: As of date of issuance and shall be implemented 
upon startup from the Unit 1 eleventh refueling and inspection outage 
currently scheduled for spring 2000.
    Amendment No.: 186.
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17029).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 30, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: December 1, 1998, as supplemented by 
your letters of April 21, July 19, October 18, and November 11, 1999.
    Brief Description of amendments: The proposed amendments would 
revise the Technical Specifications to reflect replacing the current 
Model 51 steam generators with Westinghouse Model

[[Page 4295]]

54F steam generators. The replacement program includes re-analyzing and 
evaluating loss-of-coolant-accident (LOCA) and non-LOCA mass and energy 
releases, containment and sub-compartment pressure and temperature 
responses, dose analyses, and the effects on nuclear steam supply and 
balance of plant systems.
    Date of issuance: December 29, 1999.
    Effective date: As of the date of issuance and shall be implemented 
prior to Unit 1 entering Mode 5 for Cycle 17 (Spring 2000) and prior to 
Unit 2 entering Mode 5 for Cycle 15 (Spring 2001).
    Amendment Nos.: Unit 1-147; Unit-238.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Improved Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56533). The supplemental letters dated October 18, and November 11, 
1999, provided clarifying information that did not change the initial 
proposed no significant hazards consideration determinations.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 29, 1999.
    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: July 22, 1998, as supplemented by 
letters dated June 16, October 21 and 27, November 17, and December 9, 
1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications to reflect the steam generator water level 
low-low trip setpoint differences between the existing Model E and the 
replacement Model Delta-94 steam generators for the reactor trip system 
and the engineered safety features actuation system instrumentation.
    Date of issuance: December 29, 1999.
    Effective date: December 29, 1999, to be implemented following 
replacement of Unit 1 Model E steam generators with Model Delta-94 
steam generators and prior to entry into Operational Mode 3.
    Amendment Nos.: Unit 1--120; Unit 2--108.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48268).
    The June 16, October 21 and 27, November 17, and December 9, 1999, 
supplements provided additional clarifying information that was within 
the scope of the original application and Federal Register notice and 
did not change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 29, 1999.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 19th day of January 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-1732 Filed 1-25-00; 8:45 am]
BILLING CODE 7590-01-P