[Federal Register Volume 66, Number 16 (Wednesday, January 24, 2001)]
[Notices]
[Pages 7667-7690]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-1987]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 2, 2001, through January 12, 2001. 
The last biweekly notice was published on January 10, 2001 (66 FR 
2010).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative 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's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By February 23, 2001, 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, located at One 
White Flint North, 11555 Rockville Pike (first Floor), Rockville, 
Maryland 20852. Publicly available records will be accessible 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

[[Page 7668]]

petitioner's interest. The petition should also identify the specific 
aspect(s) of the subject matter of the proceeding as to which 
petitioner wishes to intervene. Any person who has filed a petition for 
leave to intervene or who has been admitted as a party may amend the 
petition without requesting leave of the Board up to 15 days prior to 
the first prehearing conference scheduled in the proceeding, but such 
an amended petition must satisfy the specificity requirements described 
above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: December 29, 2000.
    Description of amendment request: The proposed amendment would 
increase the Technical Specification allowed outage time from 3 days to 
14 days for a single inoperable Division 1 or 2 diesel generator.
    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 any accident previously 
evaluated.
    The proposed Technical Specification (TS) changes revise the 
Completion Time for Required Actions A.2 and B.4 associated with the 
Division 1 and Division 2 Diesel Generators (DG). The proposed 
changes allow an extension of the current TS Completion Time from 72 
hours to 14 days when the Division 1 or Division 2 DG is inoperable.
    The proposed changes do not affect the design of the DGs, the 
operational characteristics of function of the DGs, the interfaces 
between the DGs and other plant systems, or the reliability of the 
DGs. Required Actions and the associated Completion Times are not 
initiating conditions for any accident previously evaluated, and the 
DGs are not initiators of any previously evaluated accidents. The 
DGs mitigate the consequences of previously evaluated accidents 
including a loss of offsite power. The consequences of a previously 
analyzed event will not be significantly affected by the extended DG 
Completion Time since the DGs will continue to be capable of 
performing their accident mitigation function as assumed in the 
accident analysis. Thus the consequences of accidents previously 
analyzed are unchanged between the existing TS requirements and the 
proposed changes. The consequences of an accident are independent of 
the time the DGs are out of service as long as adequate DG 
availability is assumed. The proposed changes will not result in a 
significant decrease in DG availability so that the assumptions 
regarding DG availability are not impacted.
    To fully evaluate the effect of the proposed EDG Completion Time 
extension, Probabilistic Risk Assessment (PRA) methods and a 
deterministic analysis were utilized. The results of the analysis 
show no significant increase in Core Damage Frequency (CDF) and 
Large Early Release Frequency (LERF). Therefore, the proposed 
changes do not involve significant increase in the probability or 
consequences of an accident previously analyzed.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a change in the design, 
configuration, or method of operation of the plant. The proposed 
changes will not alter the manner in which equipment operation is 
initiated, nor will the function demands on credited equipment be 
changed. The changes do not alter assumptions made in the safety 
analysis. No alteration in the procedures, which ensure that the 
plant remains within analyzed limits, is being proposed, and no 
changes are being made to the procedures relied upon to respond to 
an off-normal event. As such, no new failure modes are being 
introduced. Therefore, these proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    Since there are no changes to the plant design and safety 
analysis, and no changes to the DG design, including any instrument 
setpoints, no margin of safety assumed in the

[[Page 7669]]

safety analysis is affected. If a margin of safety is ascribed to DG 
availability and plant risk, it has also been determined that such a 
margin of safety is not significantly reduced, as the proposed 
changes have been evaluated both deterministically and using a risk-
informed approach. The evaluation concluded the following with 
respect to the proposed changes.
    Applicable regulatory requirements will continue to be met, 
adequate defense-in-depth will be maintained, sufficient safety 
margins will be maintained, and any increases in CDF and LERF are 
small and consistent with the NRC Safety Goal Policy Statement 
(Federal Register, Vol. 51, p. 30028 (51 FR 30028), August 4, 1986, 
as interpreted by NRC Regulatory Guides 1.174 and 1.177). 
Furthermore, increases in risk posed by potential combinations of 
equipment out of service during the proposed DG extended Completions 
Time will be managed under a configuration risk management program 
consistent with 10 CFR 50.65, ``Requirements for Monitoring the 
Effectiveness of Maintenance at Nuclear Power Plants,'' paragraph 
(a)(4). The following are examples.
     An extended DG Completion Time will not be entered 
intentionally for scheduled maintenance purposes if severe weather 
conditions are expected.
     While in the extended DG Completion Time, additional 
elective equipment maintenance or testing or equipment failure will 
be evaluated. Activities that yield unacceptable results will be 
avoided.
     The condition of the offsite power supply and 
switchyard will be evaluated.
     Activities have been identified that can mitigate any 
increase in risk. Procedures are in place for the minimizing risk 
associated with the following activities:
    No elective maintenance will be scheduled within the switchyard 
that would challenge the offsite power connection or offsite power 
availability during the extended DG Completion Time.
    No elective work will be performed on protected equipment or 
opposite train emergency core cooling system (ECCS) equipment during 
the extended DG Completion Time.
    The availability of offsite power coupled with the availability 
of the other DGs and the use of on-lime risk assessment tools 
provide adequate compensation for the potential small incremental 
increase in plant risk of the extended DG Completion Time. In 
addition, the increased availability of the DGs during refueling 
outages offsets the small increase in plant risk during operation. 
The proposed extended DG Completion Times in conjunction with the 
availability of the other DGs continues to provide adequate 
assurance of the capability to provide power to the engineered 
safety features (ESF) buses. Therefore, implementation of the 
proposed changes will not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW, Washington, DC 20036-5869.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: December 6, 2000.
    Description of amendment request: The proposed amendment requests 
changes to the once-through steam generator tube inspection criteria in 
order to allow certain inside diameter inter-granular attack 
indications to remain in service. This amendment request seeks to make 
permanent the tube inspection criteria that have been used for the past 
two operating cycles at TMI-1.
    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. Operation of the facility in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed flaw disposition strategy, based on measurable eddy 
current parameters of axial and circumferential extent for Inside 
Diameter (ID) Initiated Inter-Granular Attack (IGA), will continue 
to provide high confidence that unacceptable flaws that do not have 
the required structural integrity to withstand a postulated MSLB 
[main steam line break] are removed from service. The axial and 
circumferential length limits for eddy current ID degradation 
indications meet Draft Regulatory Guide 1.121 (Reference 9 [of the 
licensee's application]) acceptance criteria for margin to failure 
for MSLB-applied differential pressure and axial tube loads. The 
capability for detection of flaws is unaffected; and the 
identification of tubes that should be repaired or removed from 
service is maintained. The operation of the OTSGs [once-through 
steam generators] or related structures, systems, or components is 
otherwise unaffected. Therefore, neither the probability nor 
consequences of a Steam Generator Tube Rupture (SGTR) is 
significantly increased either during normal operation or due to 
limiting loads of a MSLB accident.
    Therefore, operation of the facility in accordance with the 
changes included in LCA [license change application] No. 291 will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    B. Operation of the facility in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because there are no hardware changes involved nor changes to any 
operating practices. These changes involve only the OTSG tube 
inservice inspection surveillance requirements, which could only 
affect the potential for OTSG primary-to-secondary leakage which has 
been analyzed and is subject to Technical Specification requirements 
not affected by these changes. The proposed changes continue to 
impose flaw length limits for ID IGA to assure tube structural and 
leakage integrity.
    Therefore, operation of the facility in accordance with the 
changes included with LCA No. 291 will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    C. Operation of the facility in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.
    The margins of safety defined in Draft Regulatory Guide 1.121 
(Reference 9 [of the licensee's application]) are retained. The 
probability of detecting degradation is unchanged since the bobbin 
coil eddy current methods will continue to be the primary means of 
initial detection and the probability of leakage from any 
indications left in service remains acceptably small. The strategy 
of dispositioning ID-initiated IGA indications will continue to 
provide a high level of confidence that tubes exceeding the 
allowable limits for tube integrity are repaired or removed from 
service.
    Therefore, operation in accordance with the changes included in 
LCA No. 291 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: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: December 6, 2000.
    Description of amendment request: The proposed amendment provides 
clarifications to the decay heat removal (DHR) Technical Specifications 
(TSs). It is intended, in part, to fulfill a

[[Page 7670]]

commitment made by the licensee to the NRC during a pre-decisional 
enforcement conference on April 23, 1999. Specifically, the proposed 
changes would: (1) Define and clarify the emergency feedwater (EFW) 
flowpath redundancy as described in the Bases; (2) provide operability 
requirements for the redundant steam supply paths to the turbine-driven 
EFW pump; (3) provide a more conservative 72-hour allowed outage time 
(AOT) with any EFW pump or flowpath inoperable; (4) provide a more 
conservative 1-hour AOT with both EFW flowpaths to a single once-
through steam generator (OTSG) inoperable or with 2 EFW pumps 
inoperable; and (5) revise and clarify EFW pump and flowpath 
operability requirements during surveillance testing. Minor 
administrative and editorial changes are also proposed. A change to the 
Bases for TS 3.5.5, ``Accident Monitoring Instrumentation,'' regarding 
the description of the pressurizer level instrument channels to reflect 
the replacement of Bailey transmitters was also included.
    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. Operation of the facility in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    This change incorporates the concept of EFW flowpath redundancy 
thoughout the TS[s], which takes into consideration the redundancy 
provided by the EFW System modifications made in the mid-1980s after 
the accident at TMI-2. This change incorporates a 72 hour required 
action time when redundant components are made inoperable. These 
changes do not result in any change to the configuration of the EFW 
System as described in the [UF]SAR [Updated Final Safety Analysis 
Report] or used in plant specific analyses. The reliability of EFW 
System components is unaffected. The 72 hour required action time 
for inoperability of redundant EFW components ensures that the EFW 
System can fulfill its safety function to provide adequate OTSG 
cooling during a design basis accident (DBA). The one hour required 
action time ensures prompt action to initiate a plant shutdown when 
the design flow capability of the EFW System cannot be assured.
    The current TS 4.9.1.2 contains EFW flowpath operability 
requirements during surveillance testing rather than requiring that 
a specific test be performed as do the other subparagraphs of TS 
4.9.1. For this reason the requirements of TS 4.9.1.2 are being 
moved to the LCO [limiting condition for operation] section in 
Chapter 3 and combined with the note following the current TS 
3.4.1.1.a(2) into a new TS 3.4.1.1.a(4) to define the EFW System 
operability requirements for EFW pumps and flowpaths during 
surveillance testing. The new specification incorporates the 
consideration of EFW flowpath redundancy consistent with HSPS [Heat 
Sink Protection System] train operability requirements and continues 
to require that compensatory measures be implemented to promptly 
restore components if EFW is needed during surveillance testing when 
more than one flowpath is made inoperable to an OTSG. The intent of 
this surveillance standard has been retained, which assures that the 
minimum number of EFW flowpaths to the OTSGs will be available with 
minimal operator action.
    This change provides further assurance that EFW System design 
basis requirements will be met and does not affect EFW System 
configuration, setpoints, or reliability. These changes will not 
affect any accident initiation sequence and do not affect off site 
dose consequences of accidents that have been analyzed.
    The editorial changes included in this LCA [license change 
application] are intended to improve the clarity, consistency, and 
reliability of the TS[s] [and] do not change the intent or 
interpretation.
    Therefore, operation of the facility in accordance with the 
changes included in LCA-286 will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    B. Operation of the facility in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    As a result of this change, no additional hardware is being 
added; and there will be no effect on EFW System design, operation 
as described in the [UF]SAR, or assumptions used in plant specific 
analyses. The requirement for three EFW Pumps and [associated] 
flowpaths to be operable for continuous plant operation is not 
affected by this change. Events involving the EFW System operation 
have been reviewed and determined to have no impact from these 
changes. The additional operability requirements for the turbine-
driven EFW Pump steam supplies, the revised LCOs [limiting condition 
for operation], and changes to define EFW flowpath redundancy 
ensures minimum EFW component operability as credited in plant 
analyses. The editorial changes included in this LCA are intended to 
improve clarity, consistency and readability of the TS[s] and Bases, 
[and] do not change the intent or interpretation.
    Therefore, operation of the facility in accordance with the 
changes included with LCA-286 will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    C. Operation of the facility in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.
    This change does not affect the EFW System design or 
instrumentation setpoints. The requirement for three operable EFW 
pumps and associated flowpaths is not affected by this change. The 
revised LCO imposes a 72 hour required action time when any EFW pump 
or redundant flowpath to either OTSG is inoperable, including 
inoperability for the purpose of conducting surveillance testing. 
The revised LCO requires that at least one flowpath to each OTSG 
must be operable or a plant shutdown is required to be initiated 
within one hour. The 8 hour action time currently allowed for pump 
inoperability during surveillance testing is also applied to 
flowpath inoperability during testing. The revised LCO continues to 
require compensatory measures during EFW testing when HSPS [heat 
sink protection system] is required to be operable and an OTSG is 
isolated, retaining the provision that EFW flowpath valves can be 
realligned promptly from their test mode to their operational 
allignment if EFW flow is needed. The revised Accident Monitoring 
Instrumentation specification is needed to reflect the revised 
flowpath definition and does not change the intent of the 
specification. The editorial changes included in this LCA are 
intended to improve the clarity, consistency, and readability of the 
TS[s] [and] do not change the intent or interpretation.
    Therefore, operation in accordance with the changes included in 
LCA-286 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: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendments request: December 1, 2000.
    Description of amendments request: The proposed amendments would 
revise the value of the minimum departure from nucleate boiling ratio 
(DNBR) from `` 1.30'' in the current technical 
specifications to `` 1.3 (through operating cycle 10)'' and 
`` 1.34 (operating cycle 11 and later)'' in the safety 
limits Technical Specification (TS) 2.1.1.1 and in function 15, DNBR--
Low, in Table 3.3.1-1, ``Reactor Protective System Instrumentation.'' 
The proposed amendments are structured such that the `` 
1.34'' would become effective for each unit in operating cycle 11 and 
later. Operating cycle 11 begins in spring 2002 for Unit

[[Page 7671]]

2, in fall 2002 for Unit 1, and in spring 2003 for Unit 3. From now to 
operating cycle 11, the `` 1.30'' will remain the minimum 
DNBR requirement for the three units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Standard 1--Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The purpose of the proposed Technical Specification (TS) 
amendment is to provide a revised Departure from Nucleate Boiling 
Ratio (DNBR) Safety Limit (TS Section 2.1.1.1) and Low DNBR Reactor 
Protective System (RPS) trip setpoint (TS Limiting Condition for 
Operation (LCO) 3.3.1, Table 3.3.1-1).
    The proposed TS amendment involves increasing the DNBR Safety 
Limit and Low DNBR RPS trip setpoint from `` 1.30'' to 
`` 1.34''. Changing this limit in and of itself will not 
alter the physical characteristics of any component involved in the 
initiation of an accident. Thus, the proposed change does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The Core Operating Limit Supervisory System (COLSS) Power 
Operating Limit (POL) is an alarm limit on the maximum steady state 
core power level. The alarm is based on maintaining COLSS calculated 
DNBR a pre-determined amount above the DNBR Safety Limit. The Low 
DNBR RPS trip setpoint[,] in conjunction with the COLSS POL, 
prevents the DNBR in the limiting coolant channel in the core from 
violating the DNBR Safety Limit during design basis Anticipated 
Operational Occurrences (AOO). Operating below the COLSS POL ensures 
the Low DNBR RPS trip setpoint will protect the core [fuel] from 
damage due to the occurrence of locally saturated conditions in the 
limiting (hot) channel during the worst AOO. Thus, during normal and 
anticipated operation the Low DNBR RPS trip setpoint in conjunction 
with the COLSS POL prevents overheating of the fuel cladding and 
subsequent cladding perforation that would release fission products 
to the reactor coolant.
    This change will accommodate increased DNBR sensitivity to 
uncertainties in inlet flow to the hot assembly and adjacent 
assemblies. This increased sensitivity is attributed to the flatter 
power distributions of the more efficient present day erbium core 
designs. More adverse DNBR sensitivity to inlet flow was first 
encountered in Unit 1 Cycle 7. At that time the increased DNBR 
sensitivity was accounted for statistically by applying a thermal 
margin penalty to Core Operating Limit Supervisory System (COLSS) 
and Core Protection Calculators (CPCs) using approved Statistical 
Combination of Uncertainties (SCU) methods. This approach was also 
used for the subsequent cycles in all units up until the present. 
The NRC Safety Evaluation (issued May 26, 1994 for Palo Verde 
Nuclear Generating Stations (PVNGS) Units 1, 2, and 3) for the 
present `` 1.30'' DNBR limit states, ``Uncertainties in 
inlet flow to the hot assembly and adjacent assemblies can be 
accounted for statistically by either increasing DNBR or applying a 
thermal margin penalty using approved SCU methods.''
    The proposed TS amendment change for DNBR Safety Limit and Low 
DNBR RPS trip setpoint limit ( 1.34) was calculated using 
approved SCU methods to statistically include the above described 
increased DNBR sensitivity. This new DNBR limit was calculated such 
that it has a high probability of covering all future cycle designs. 
Thus, this change involves moving the existing increased inlet flow 
uncertainty penalty from a thermal margin penalty contained within 
COLSS and CPCs to an increase in the DNBR Safety Limit and Low DNBR 
RPS trip setpoint limit. The DNBR Safety Limit and Low DNBR RPS trip 
setpoint increases from `` 1.30'' to `` 1.34'' 
due to this change. The COLSS and CPCs would respond similarly with 
the increased inlet flow uncertainty penalty located in either the 
COLSS or CPCs or in the DNBR Safety Limit. The proposed amendment 
changes only the location of the increased inlet flow uncertainty 
penalty and does not impact the operation of the plant. The core 
power distribution during all phases of normal and anticipated 
operational occurrences will remain bounded by the initial 
conditions assumed in Chapter 15 of the Palo Verde Nuclear 
Generating Station (PVNGS) UFSAR [Updated Final Safety Analysis 
Report]. Thus, the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    Standard 2--Does the proposed change create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    No. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This change does not alter the physical design of any System, 
Structure, or Component (SSC) of the plant.
    The change involves increasing the DNBR Safety Limit and the Low 
DNBR RPS trip setpoint from `` 1.30'' to `` 
1.34'' and decreasing the corresponding DNBR thermal margin penalty 
factors in COLSS and CPC in a compensating manner. Changing these 
limits and penalty factors will not alter the physical or functional 
characteristics of any component in the plant. These changes will 
not affect any safety-related equipment used in the mitigation of 
anticipated operational occurrences or design basis accidents. Thus, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Standard 3--Does the proposed change involve a significant 
reduction in a margin of safety?
    No. The proposed change does not involve a significant reduction 
in a margin of safety.
    The DNBR Safety Limit specified in Section 2.1.1.1 and the Low 
DNBR RPS trip setpoint specified in Table 3.3.1-1 of LCO 3.3.1 of 
[the] PVNGS Technical Specifications ensure that operation of the 
reactor does not result in a departure from nucleate boiling during 
normal operation and design basis anticipated operational 
occurrences. Therefore, operating consistent with the increased DNBR 
Safety Limit and Low DNBR RPS trip setpoint will ensure that no 
anticipated operational occurrences will result in core conditions 
below the specified DNBR Safety Limit and no postulated accident 
exceeds the site boundary dose limits. The UFSAR Chapter 15 analysis 
remains bounding and the margins of safety will be maintained 
because the COLSS and the CPC overall uncertainty factors will be 
calculated and implemented consistent with the increased DNBR Safety 
Limit of `` 1.34''. Therefore, this change to TS Section 
2.1.1.1 and Table 3.3.1-1 of LCO 3.3.1 does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendments request: December 5, 2000.
    Description of amendments request: The proposed amendments would 
revise the action statement for Specification 3.7.5, ``Auxiliary 
Feedwater (AFW) System,'' of the Technical Specifications (TSs). The 
amendments would incorporate NRC-approved TS Task Force (TSTF) Traveler 
Number TSTF-340, Revision 3, to allow a 7-day Completion Time for the 
turbine-driven AFW pump if inoperability occurs in reactor Mode 3 
following a refueling outage, and if Mode 2 had not been entered.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or

[[Page 7672]]

consequences of an accident previously evaluated?
    No. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment to Technical Specification 3.7.5 would 
allow a 7 day Completion Time for Condition A for the turbine-driven 
Auxiliary Feedwater (AFW) pump if the inoperability occurs in MODE 3 
following a refueling outage, if MODE 2 had not been entered. 
Extending the Completion Time does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because: (1) The proposed amendment does not 
represent a change to the system design, (2) the proposed amendment 
does not prevent the safety function of the AFW system from being 
performed since the other fully redundant esstential train and the 
non-essential train are required to be operable, (3) the proposed 
amendment does not alter, degrade, or prevent action described or 
assumed in any accident described in the PVNGS [Palo Verde Nuclear 
Generating Station] UFSAR [Updated Final Safety Analysis Report] 
from being performed since the other trains of AFW are required to 
be operable, (4) the proposed amendment does not alter any 
assumptions previously made in evaluating radiological consequences, 
and (5) the proposed amendment does not affect the integrity of any 
fission product barrier. No other safety related equipment is 
affected by the proposed change. Therefore, this proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to Technical Specification 3.7.5 would 
allow a 7 day Completion Time for Condition A for the turbine-driven 
Auxiliary Feedwater (AFW) pump if the inoperability occurs in MODE 3 
following a refueling outage, if MODE 2 had not been entered. 
Extending the Completion Time does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated because: (1) The proposed amendment does not represent a 
change to the system design, (2) the proposed amendment does not 
alter how equipment is operated or the ability of the system to 
deliver the required AFW flow, and (3) the proposed amendment does 
not affect any other safety related equipment. Therefore, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The PVNGS safety analysis credits essential Auxiliary Feedwater 
(AFW) pump delivery of 650 gpm at a steam generator pressure of 1270 
psia or equivalent at the steam generator entrance for design basis 
accidents. The AFW System Design Basis Manual (AF), Revision 11, 
states that these pumps are designed to supply 750 gpm. The proposed 
[***] amendment to Technical Specification 3.7.5 would allow a 7 day 
Completion Time for Condition A for the turbine-driven AFW pump if 
the inoperability occurs in MODE 3 following a refueling outage, if 
MODE 2 had not been entered. Extending the Completion Time does not 
involve a significant reduction in a margin of safety because: (1) 
During a return to power operations following a refueling outage, 
decay heat [in the core] is at its lowest levels, (2) the other 
essential and non-essential AFW trains are required to be OPERABLE 
when MODE 3 is entered, (3) the essential motor-driven AFW train can 
provide sufficient flow to remove decay heat and cool the unit to 
Shutdown Cooling system entry conditions from power operations, and 
4) the non-essential motor-driven AFW train is designed to supply 
sufficient water to remove decay heat with steam generator pressure 
at no load conditions to cool the unit to Shutdown Cooling entry 
conditions.
    Based on the responses to these three criteria, APS [Arizona 
Public Service Company] has concluded that the proposed amendment 
involves no significant hazards considerations.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: December 13, 2000
    Description of amendment request: The proposed amendment would 
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.9.2 
``Refueling Operations--Instrumentation'' and the associated Bases to 
permit using alternate installed detectors or temporary source range 
detectors instead of the two Source Range Nuclear Flux Monitors 
specified in the current HNP TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change only involves reactor core monitoring requirements 
during Mode 6. These monitoring requirements are not credited for 
accident mitigation. Alternate monitors will be provided with the 
accuracy and sensitivity required to adequately monitor changes in 
the core reactivity levels during refueling activities. Neutron Flux 
monitors are for indication only and do not interface with other 
structures, systems, or components that might initiate an accident.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Neutron Flux monitors are for indication only and do not 
interface with other structures, systems, or components that might 
initiate an accident. The proposed change will not modify plant 
systems or operate plant components such that a new or different 
accident scenario 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 amendment does not involve a significant 
reduction in the margin of safety.
    Similar changes, to the proposed change, have been approved at 
the Beaver Valley Power Station and the Diablo Canyon Power Plant. 
The proposed change will maintain adequate monitoring of core 
reactivity in Mode 6. The proposed change maintains requirements for 
two operable neutron flux monitors. Neutron flux monitors are not 
credited in the HNP accident analyses for accident mitigation in 
Mode 6.
    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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: December 14, 2000.

[[Page 7673]]

    Description of amendment request: The proposed amendment would 
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.8.1 
related to emergency diesel generators (EDGs). Specifically, the 
licensee proposes revising TS Surveillance Requirement 4.8.1.1.2.f.7, 
the 24-hour EDG endurance run test, by removing the restriction to 
perform the test during shutdown conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The EDGs and their associated emergency buses are not accident 
initiating equipment; therefore, there will be no impact on accident 
probabilities due to this proposed amendment. The EDGs mitigate the 
consequences of previously evaluated accidents involving a loss of 
offsite power. The proposed amendment continues to assure the EDGs 
perform their function when called upon. The design of the equipment 
is not being modified. The proposed amendment does not impact the 
operational characteristics of the EDGs, the interfaces between the 
EDGs and other plant systems, or the function or reliability of the 
EDGs. The EDGs remain capable of performing their accident 
mitigation function. The HNP Probabilistic Safety Analysis (PSA) 
model results are not affected by the proposed change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not alter the design, configuration, 
or method of operation of the plant. No physical changes are being 
proposed, nor any changes to the method of operation of the EDGs or 
supporting systems. The proposed amendment, in effect, allows a 
small increase in the duration that the EDGs are operated parallel 
to the grid for test purposes. No new system interactions are 
created, and the proposed change does not introduce a new failure 
mode.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed change does not affect the Limiting Conditions for 
Operation or their Bases that are used to establish any margin of 
safety. The ability of the EDGs to separate from the offsite power 
source has been designed and tested per Technical Specification 
requirements. The proposed change does not involve a change to the 
plant design or operation and does not affect the availability of 
any of the required power sources, nor the capability of the EDGs to 
perform their intended safety function.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The foregoing analysis demonstrates that the proposed amendment 
to HNP TS does not involve a significant increase in the probability 
or consequences of a previously evaluated accident, does not create 
the possibility of a new or different kind of accident, and does not 
involve a significant reduction in a margin of safety.
    Based upon the preceding analysis, [Carolina Power & Light 
Company] CP&L concludes that the proposed amendment 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: December 11, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 3.1.F.2.a, ``Primary to Secondary 
Leakage,'' and 4.13.A.3.f, ``Steam Generator Tube Inservice 
Surveillance,'' based on the prior replacement of the steam generators 
(SGs). Specifically, the proposed changes would (1) revise the primary 
to secondary leakage limits and (2) delete requirements associated with 
tube sleeve repair, steam generator tube denting, F* repair 
classification and criteria, and (3) modify the associated TS Bases. In 
addition, the proposed amendment includes several related 
administrative 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:

Changes to SG Primary to Secondary Leakage Limits

    1. Does the change involve a significant increase in the 
probability [* * *] or consequences of an accident previously 
evaluated?
    The proposed reduction in primary to secondary leakage limit and 
the elimination of the limit for SGs containing sleeved tubes does 
not affect accident initiators or precursors. The proposed change 
establishes a primary to secondary leakage limit that is equivalent 
to the lesser of the primary to secondary leakage limits currently 
established for SG with and without SG tube sleeves. Reducing the 
primary to secondary leakage limit does not increase the probability 
of an accident. The proposed change does not increase primary to 
secondary leakage limits. Therefore, the consequences of an accident 
are not increased. Therefore, the probability of occurrence or the 
consequences of accidents previously evaluated are not significantly 
increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not modify any plant equipment. 
Therefore, the proposed changes do not degrade the reliability of 
systems, structures, or components or create a new accident 
initiator or precursor. No new failure modes are created. Therefore, 
the change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change establishes one limit for primary to 
secondary limit that is the same as the most restrictive of the two 
primary to secondary leakage limits that currently exists. The 
proposed change does not increase the allowable primary to secondary 
leakage limit.
    Since the primary to secondary leakage limit is not increased, 
the margin of safety will not be reduced. The proposed change still 
requires verification that primary to secondary leakage is within 
the limit at the existing frequency. Since the primary to secondary 
leakage limit is not increased, dose rates at the site boundary will 
not be increased. Therefore, the proposed activity does not involve 
a significant reduction in a margin of safety.

Deletion of Provisions Associated With SG Tube Sleeving Repair 
Method

    1. Does the change involve a significant increase in the 
probability [* * *] or consequences of an accident previously 
evaluated?
    The proposed deletion of the SG tube sleeving provisions does 
not affect accident initiators or precursors. The proposed change 
deletes the TS provisions that are not approved for the replacement 
SGs. Deletion of an unapproved repair method from the TS does not 
increase the probability of an accident and the proposed change does 
not increase primary to secondary leakage limits. Consequently, the 
consequences of an accident are not significantly increased. 
Therefore, the probability of occurrence or

[[Page 7674]]

the consequences of accidents previously evaluated are not 
significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not impact or interface with plant 
safety related equipment. Therefore, the proposed changes do not 
degrade the reliability of systems, structures, or components or 
create a new accident initiator or precursor. No new failure modes 
are created. Therefore, the change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change deletes the TS provisions that are not 
approved for the replacement SGs. The proposed change does not 
increase the allowable primary to secondary leakage limit. Since the 
primary to secondary leakage limit is not increased, the margin of 
safety will not be reduced. Therefore, the proposed activity does 
not involve a significant reduction in a margin of safety.

Deletion of Provisions Associated with Steam Generator F* Tube 
Classification

    1. Does the change involve a significant increase in the 
probability [* * *] or consequences of an accident previously 
evaluated?
    The proposed deletion of the F* criteria and associated 
provisions does not affect accident initiators or precursors. The 
proposed change deletes the TS provisions that are not approved for 
the replacement SGs. Deletion of an unapproved repair method from 
the TS does not increase the probability of an accident. The 
proposed change does not increase primary to secondary leakage 
limits. Therefore, the probability of occurrence or the consequences 
of accidents previously evaluated are not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not impact or interface with plant 
safety related equipment. Therefore, the proposed changes do not 
degrade the reliability of systems, structures, or components or 
create a new accident initiator or precursor. No new failure modes 
are created. Therefore, the change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change deletes the TS provisions that are not 
approved for the replacement SGs. The proposed change does not 
increase the allowable primary to secondary leakage limit. Since the 
primary to secondary leakage limit is not increased, the margin of 
safety will not be reduced. Therefore, the proposed activity does 
not involve a significant reduction in a margin of safety.

Deletion of Provisions Associated With SG Tube Denting Phenomenon

    1. Does the change involve a significant increase in the 
probability [* * *] or consequences of an accident previously 
evaluated?
    The proposed deletion of the requirements and associated 
provisions regarding SG tube denting does not significantly affect 
accident initiators or precursors. The proposed change deletes from 
the TS provisions that are not necessary for the replacement SGs. 
Deletion of the SG tube denting examination requirements from the TS 
does not increase the probability of an accident. The proposed 
change does not increase primary to secondary limits. Therefore, the 
consequences of an accident are not increased. Therefore, the 
probability of occurrence or the consequences of accidents 
previously evaluated are not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not impact or interface with plant 
safety related equipment. Therefore, the proposed changes do not 
degrade the reliability of systems, structures, or components or 
create a new accident initiator or precursor. No new failure modes 
are created. Therefore, the change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change deletes the TS provisions that are not 
applicable for the replacement SGs. The proposed change does not 
increase the allowable primary to secondary leakage limit. Since the 
primary to secondary leakage limit is not increased, the margin of 
safety will not be reduced. Therefore, the proposed activity does 
not involve a significant reduction in a margin of safety.

Related Administrative Changes

    1. Does the change involve a significant increase in the 
probability [* * *] or consequences of an accident previously 
evaluated?
    The proposed administrative changes do not affect accident 
initiators or precursors. The proposed changes correct the 
presentation of several TS Basis pages and delete an obsolete 
scheduler extension footnote. Correcting the page presentation and 
deleting an obsolete footnote do not increase the probability of an 
accident. The proposed change does not increase primary to secondary 
leakage limits. Consequently, the consequences of an accident are 
not significantly increased.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not impact or interface with plant 
safety related equipment. Therefore, the proposed changes do not 
degrade the reliability of systems, structures, or components or 
create a new accident initiator or precursor. No new failure modes 
are created. Therefore, the change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed administrative changes do not affect accident 
initiators or precursors. The proposed change corrects the 
presentation of several TS Basis pages and deletes an obsolete 
scheduler extension footnote. The proposed changes do not increase 
the allowable primary to secondary leakage limit. Since the primary 
to secondary leakage limit is not increased, the margin of safety 
will not be reduced. Therefore, the proposed activity 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Marsha Gamberoni.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: December 7, 2000.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) regarding the Limiting 
Conditions for Operation (LCO) for the auxiliary feedwater system (LCO 
3.7.5) to be similar to changes to the ``Standard Technical 
Specifications, Combustion Engineering Plants,'' NUREG 1432, Revision 1 
(STS), made by the Nuclear Energy Institute Technical Specifications 
Task Force (TSTF) change number 325, ``Changes To Structure Of 
[Emergency Core Cooling System] ECCS--Operating LCO.''
    Palisades LCO 3.7.5, ``Auxiliary Feedwater System,'' would be 
changed as follows: (1) An editorial change would be made to Note 2 to 
put the word ``operable'' in uppercase letters; (2) the second and 
third parts of the Condition A description, ``AND--At least 100% of the 
required AFW flow available to each steam generator--AND--At least two 
AFW pumps OPERABLE,'' would be deleted; (3) the second part of the 
Condition B description, ``One or more AFT trains inoperable for 
reasons other than Condition A with at least 100% of the required AFW 
flow available in MODE 1, 2, or 3,'' would be replaced with two new 
parts (``Less than 100% of the required AFW flow available to either 
steam generator--OR--Fewer than two AFW pumps OPERABLE in mode 1, 2, OR 
3''); and (4) the wording of

[[Page 7675]]

Condition C would be revised to address the condition where 
insufficient AFW flow is available to achieve a plant shutdown while in 
any mode within the applicable conditions of LCO 3.7.5. The licensee 
also forwarded related changes to the TS Bases.
    Additional changes requested in the licensee's application dated 
December 7, 2000, are based upon other TSTFs and are addressed by 
separate Federal Register notices.
    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 changes would not] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Changes are proposed to LCO 3.7.5, Auxiliary Feedwater, which 
emulate changes made to Standard Technical Specifications, 
Combustion Engineering Plants, NUREG 1432, Rev.1 (STS) by TSTF 325. 
The structure of LCO 3.7.5 has been rearranged to maintain Condition 
A (and, in certain circumstances, Condition B) in effect if failures 
should occur which reduce available flow to less than 100% of the 
required flow (that flow assumed in the accident analyses). The 
resulting requirements are those intended when the LCO was initially 
constructed and represent the way the LCO Conditions are being 
applied. Therefore there is no change in intent or application of 
the LCO. In the case where inoperable AFW train components reduce 
available flow below that required, and a subsequent partial 
restoration is made to provide 100% of the required flow, the 
proposed change makes the literal requirements more conservative 
because (with the proposed arrangement) the Completion Time for 
Condition A (and possibly Condition B) would start when the initial 
inoperability occurred rather than (with literal interpretation of 
the existing arrangement) when Condition A (or B) was entered after 
the partial restoration. . . .
    As described above, the proposed change corrects the structure 
of the LCO to assure its correct application. There is no change in 
intent or in the way the LCO is actually applied. The literal (and 
unintended) interpretation of the existing LCO structure could, 
under some circumstances, provide longer than intended Completion 
Times for restoration of operability. The proposed change only 
clarifies the requirements of the LCO Required Actions. Since the 
proposed change affects neither the LCO intent nor its application, 
the proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    B. [The proposed changes would not] create the possibility of a 
new or different kind of accident from any previously evaluated.
    As described above, the proposed change corrects the structure 
of the LCO to assure its correct application. There is no change in 
intent or in the way the LCO is actually applied. The proposed 
changes would not result in any physical alterations to the plant 
configuration, no new equipment is added, no equipment interfaces 
are modified, no changes to any equipment's function or the method 
of operating the equipment are being made. As the proposed changes 
would not change the design, configuration or operation of the 
plant, no new or different kinds of accident modes are created. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    C. [The proposed changes would not] involve a significant 
reduction in a margin of safety.
    As described above, the proposed change corrects the structure 
of the LCO to assure its correct application. The proposed changes 
are consistent with the intent of the changes made to the STS by 
TSTF 325. There is no change in intent or in the way the LCO is 
actually applied. 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: December 7, 2000.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) in accordance with changes to 
the ``Standard Technical Specifications, Combustion Engineering 
Plants,'' NUREG 1432, Revision 1 (STS), made by the Nuclear Energy 
Institute Technical Specifications Task Force (TSTF) change number 325, 
``Changes To Structure Of [Emergency Core Cooling System] ECCS--
Operating [Limiting Condition for Operation] LCO.'' Specifically, 
Palisades LCO 3.5.2, ``ECCS--Operating,'' would be changed as follows: 
(1) the second part of the Condition B description, ``At least 100% of 
the required ECCS flow available,'' would be deleted; (2) the wording 
of Condition C would be revised to limit its application to Conditions 
A or B; and (3) the wording that would be removed from Condition B 
would be made into a new condition, Condition D, which would read: 
``Less than 100% of the required ECCS flow available.'' Required Action 
D.1, ``Enter LCO 3.0.3,'' and its completion time, ``Immediately,'' 
would also be added. The licensee also forwarded related changes to the 
TS Bases.
    Additional changes requested in the licensee's application dated 
December 7, 2000, are based upon other TSTFs and are addressed by 
separate Federal Register notices.
    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 change is proposed which emulates changes made to Standard 
Technical Specifications, Combustion Engineering Plants, NUREG 1432, 
Rev. 1 (STS) by TSTF 325. The structure of LCO 3.5.2, ECCS--
Operating, has been rearranged to maintain Condition B in effect if 
failures should occur which reduce available flow to less than 100% 
of the required flow (that flow assumed in the accident analyses). 
The resulting requirements are those intended when the LCO was 
initially constructed and represent the way the LCO Conditions are 
being applied. Therefore there is no change in intent or application 
of the LCO. In the case where inoperable ECCS train components 
reduce available flow below that required, and a subsequent partial 
restoration is made to provide 100% of the required flow, the 
proposed change makes the literal requirements more conservative 
because (with the proposed arrangement) the Completion Time for 
Condition B would start when the initial inoperability occurred 
rather than (with literal interpretation of the existing 
arrangement) when Condition B was entered after the partial 
restoration. * * *
    A. [The proposed changes would not] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    As described above, the proposed change corrects the structure 
of the LCO to assure its correct application. There is no change in 
intent or in the way the LCO is actually applied. The literal (and 
unintended) interpretation of the existing LCO structure could, 
under some circumstances, provide longer than intended Completion 
Times for restoration of operability. The proposed change only 
clarifies the requirements of the LCO Required Actions. Since the 
proposed change affects neither the LCO intent nor its application, 
the proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    B. [The proposed changes would not] create the possibility of a 
new or different kind of accident from any previously evaluated.
    As described above, the proposed change corrects the structure 
of the LCO to assure its correct application. There is no change in 
intent or in the way the LCO is actually applied. The proposed 
changes would not

[[Page 7676]]

result in any physical alterations to the plant configuration, no 
new equipment is added, no equipment interfaces are modified, and no 
changes to any equipment's function or the method of operating the 
equipment are being made. As the proposed changes would not change 
the design, configuration or operation of the plant, no new or 
different kinds of accident modes are created. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    C. [The proposed changes would not] involve a significant 
reduction in a margin of safety.
    As described above, the proposed change corrects the structure 
of the LCO to assure its correct application. The proposed change is 
consistent with the requirements of the STS. There is no change in 
intent or in the way the LCO is actually applied. 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: December 7, 2000 (this application 
supercedes an amendment request dated July 28, 2000).
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to allow Type B and C 
containment leak rate testing to be performed in accordance with 10 CFR 
part 50, appendix J, option B. Conversion to Option B affects TS 5.5.14 
and Surveillance Requirements (SRs) SR 3.6.1.1, SR 3.6.1.3, and SR 
3.6.2.1. The proposed amendment also revises the SR 3.6.2.2 frequency 
for containment air lock door interlock testing from 18 months to 24 
months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    * * * Four groups of changes have been proposed:
    First, changes are proposed to allow Type B and C containment 
leak rate testing to be performed in accordance with 10 CFR 50, 
appendix J, Option B.
    Second, exceptions are proposed to the Option B testing 
methodology for containment air lock door seals.
    Third, an exception is proposed to the Option B testing 
frequency for small diameter containment purge valves.
    Fourth, the frequency for the containment air lock door 
interlock testing has been extended from 18 months to 24 months.
    The following evaluation supports the finding that operation of 
the facility in accordance with the proposed changes would not:
    a. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    All four groups of proposed changes deal exclusively with 
testing of features related to containment isolation. The changes 
only affect testing frequency and methodology. The proposed testing 
methodologies are acceptable under the existing Technical 
Specifications. None of the devices involved are assumed as an 
initiator of any accident previously evaluated. Therefore, operation 
of the facility in accordance with the proposed changes would not 
involve a significant increase in the probability of an accident.
    1. The first group of proposed changes is based on the model 
Technical Specifications approved by the NRC staff in TSTF 
[Technical Specification Task Force] 52, Rev. 3. Test intervals will 
be established based on performance history of the components 
tested. The frequency of testing the containment penetrations and 
containment isolation valves will be extended in accordance with 
program requirements and 10 CFR 50, appendix J, Option B, with 
reference to Regulatory Guide 1.163, and NEI [Nuclear Energy 
Institute] 94-01, Rev 0. The change in risk resulting from the 
proposed changes was evaluated by the NRC in the rule making process 
for implementing the Option B requirements and are characterized in 
NUREG-1493. For Type B and C tests the NRC concluded that the 
extension of test intervals as allowed by Option B would lead to 
only minor increases in potential offsite dose consequences. These 
increases are offset by the expected decrease in worker dose 
received during Type A, B, and C testing, and were found to be 
acceptable. Therefore, operation of the facility in accordance with 
the first group [of] proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The second group of proposed changes would allow air lock 
door seal leak rate testing to be performed by a seal contact check 
(for the Emergency Escape Air Lock) or by pressurizing between the 
door seals at a pressure [greater than or equal to] 10 psig (for the 
Personnel Air Lock) following door seal contact adjustments. Both 
proposed alternative testing methods are allowed by existing 
Technical Specifications (while testing under Option A) and both 
will result in a continuation of the currently successful testing 
practice which has provided a high degree of confidence in door seal 
performance. Plant operating history has shown that air lock door 
seals which have been successfully tested in accordance with the 
proposed methodology have passed subsequent full pressure air lock 
leakage tests in virtually every case.
    Since the proposed methodology has been demonstrated to 
successfully detect leaking door seals, the continued use of that 
methodology for testing under the requirements of Option B will not 
cause an increase in the probability of a leaking air lock door seal 
going undetected. Also, since there will be no increase in the rate 
of occurance [sic] of undetected leakage due to the continued 
utilization of current practices under Option B, operation of the 
facility in accordance with the second group of proposed changes 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    3. The third proposed change allows the testing frequency for 
the Containment 4-inch purge exhaust, 8-inch purge exhaust and 12-
inch air room supply valves to be consistent with other 10 CFR 50, 
appendix J, Option B, Type C test intervals and is supported by 
Palisades design, historical test results and other required 
testing. This would allow the test interval to be extended to a 
maximum of 60 months from the 30 month interval allowed without this 
exception.
    The change in risk resulting from the third proposed change is 
essentially the same as that evaluated by the NRC in the rule making 
process for implementing the Option B Type C testing requirements, 
which are characterized in NUREG-1493. As discussed under change 1, 
above, the NRC concluded that the extension of test intervals as 
allowed by Option B for Type C testing would lead to only minor 
increases in potential offsite dose consequences. These increases 
were found to be acceptable. The third proposed change applies this 
longer interval to moderate diameter valves in the containment purge 
system. That longer interval would apply to these valves, without 
the proposed exception, if they were installed as containment 
isolation valves in a different system. Furthermore, the 8-inch and 
12-inch valves are effectively leak rate tested on a 184 day 
frequency as part of their required closure verification. Therefore, 
the proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    4. The fourth proposed change only extends the frequency for 
containment air lock door interlock testing. The proposed change 
will not affect any parameters or conditions that contribute to the 
mitigation of previously evaluated accidents. Therefore, operation 
of the facility in accordance with the fourth proposed change would 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    b. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    All four groups of proposed changes deal exclusively with 
testing of features related to containment isolation. The changes 
only affect testing frequency and methodology. The proposed testing 
methodologies are acceptable under the existing Technical 
Specifications. The proposed changes would not result in any 
physical alterations to the

[[Page 7677]]

plant configuration, no new equipment is added, no equipment 
interfaces are modified, no changes to any equipment's function or 
the method of operating the equipment are being made. As the 
proposed changes would not change the design, configuration or 
operation of the plant, they would not cause the containment leak 
rate testing to become an accident initiator. No new or different 
kinds of accident modes are created. Therefore, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any previously evaluated.
    c. Involve a significant reduction in the margin of safety.
    All four groups of proposed changes deal exclusively with 
testing of features related to containment isolation. The changes 
only affect testing frequency and methodology. The proposed testing 
methodologies are acceptable under the existing Technical 
Specifications. None of the devices involved are assumed as an 
initiator of any accident previously evaluated. The proposed changes 
only affect the methodology and frequency of Type B and C testing. 
The methods for performing the tests are not changed from those 
specified in existing Technical Specifications. The proposed 
performance based approach, provided by using Option B to 10 CFR 50, 
Appendix J, would continue to ensure that the containment leakage 
rates would not exceed the maximum allowable leakage rates defined 
in the Technical Specifications and assumed in the 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: December 7, 2000.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) regarding the Limiting 
Conditions for Operation (LCO) for the containment cooling systems (LCO 
3.6.6), the component cooling water system (LCO 3.7.7), and the service 
water system (LCO 3.7.8) to be similar to changes to the ``Standard 
Technical Specifications, Combustion Engineering Plants,'' NUREG 1432, 
Revision 1 (STS), made by the Nuclear Energy Institute Technical 
Specifications Task Force (TSTF) change number 325, ``Changes To 
Structure Of [Emergency Core Cooling System] ECCS--Operating LCO.''
    Palisades LCO 3.6.6, ``Containment Cooling Systems,'' would be 
changed as follows: (1) The second part of the Condition A description, 
``AND--At least 100% of the required post accident containment cooling 
capability available,'' would be deleted; (2) the wording of Condition 
B would be revised to limit its application to Condition A; and (3) the 
wording removed from Condition A would be made into a new condition, 
Condition C, which would read: ``Less than 100% of the required post-
accident containment cooling capability available.'' Required Action 
C.1, ``Enter LCO 3.0.3,'' and its completion time, ``Immediately,'' 
would also be added. The licensee also forwarded related changes to the 
TS Bases.
    Palisades LCO 3.7.7, ``Component Cooling Water [CCW] System,'' 
would be changed as follows: (1) The second part of the Condition A 
description, ``AND--At least 100% of the required CCW post accident 
capability available,'' would be deleted; (2) the wording of Condition 
B would be revised to limit its application to Condition A; and (3) the 
wording removed from Condition A would be made into a new condition, 
Condition C, which would read: ``Less than 100% of the required post-
accident CCW capability available.'' Required Action C.1, ``Enter LCO 
3.0.3,'' and its completion time, ``Immediately,'' would also be added. 
The licensee also forwarded related changes to the TS Bases.
    Palisades LCO 3.7.8, ``Service Water System [SWS],'' would be 
changed as follows: (1) The second part of the Condition A description, 
``AND--At least 100% of the required post accident SWS capability 
available,'' would be deleted; (2) the wording of Condition B would be 
revised to limit its application to Condition A; and (3) the wording 
removed from Condition A would be made into a new condition, Condition 
C, which would read: ``Less than 100% of the required post-accident SWS 
capability available.'' Required Action C.1, ``Enter LCO 3.0.3,'' and 
its completion time, ``Immediately,'' would also be added. The licensee 
also forwarded related changes to the TS Bases.
    Additional changes requested in the licensee's application dated 
December 7, 2000, are based upon other TSTFs and are addressed by 
separate Federal Register notices.
    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:

    Changes are proposed for three Palisades LCOs structured like 
LCO 3.5.2 which emulate changes made to Standard Technical 
Specifications, Combustion Engineering Plants, NUREG 1432, Rev.1 
(STS) by TSTF 325. The structure of LCOs 3.6.6, 3.7.7, and 3.7.8 has 
been rearranged to maintain Condition A in effect if failures should 
occur which reduce available flow to less than 100% of the required 
cooling capability (that assumed in the accident analyses). The 
resulting requirements are those intended when the LCOs were 
initially constructed and represent the way the LCO Conditions are 
being applied. Therefore there is no change in intent or application 
of the LCOs. In the case where inoperable required components reduce 
available cooling below that required, and a subsequent partial 
restoration is made to provide 100% of the required cooling, the 
proposed change makes the literal requirements more conservative 
because (with the proposed arrangement) the Completion Time for 
Condition A would start when the initial inoperability occurred 
rather than (with literal interpretation of the existing 
arrangement) when Condition A was entered after the partial 
restoration. * * *
    A. [The proposed changes would not] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    As described above, the proposed changes correct the structure 
of the subject LCOs to assure their correct application. There is no 
change in intent or in the way the LCOs are actually applied. The 
literal (and unintended) interpretation of the existing LCO 
structure could, under some circumstances, provide longer than 
intended Completion Times for restoration of operability. The 
proposed changes only clarify the requirements of the LCO Required 
Actions. Since the proposed changes affect neither the LCO intent 
nor their application, the proposed changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. [The proposed changes would not] create the possibility of a 
new or different kind of accident from any previously evaluated.
    As described above, the proposed changes correct the structure 
of LCOs 3.6.6, 3.7.7, and 3.7.8 to assure their correct application. 
There is no change in intent or in the way the LCOs are actually 
applied. The proposed changes would not result in any physical 
alterations to the plant configuration, no new equipment is added, 
no equipment interfaces are modified, and no changes to any 
equipment's function or the method of operating the equipment are 
being made. As the proposed changes would not change the design, 
configuration or operation of the plant, no new or different kinds 
of accident modes are created. Therefore, the proposed

[[Page 7678]]

changes do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    C. [The proposed changes would not] involve a significant 
reduction in a margin of safety
    As described above, the proposed changes correct the structure 
of the subject LCOs to assure their correct application. The 
proposed changes are consistent with the changes made to the STS by 
TSTF 325. There is no change in intent or in the way the LCOs are 
actually applied. 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: December 7, 2000.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) in accordance with changes to 
the ``Standard Technical Specifications, Combustion Engineering 
Plants,'' NUREG 1432, Revision 1, made by the Nuclear Energy Institute 
Technical Specifications Task Force (TSTF) change number 258, Revision 
4. TSTF 258 addresses changes to various Administrative Controls TSs. 
The licensee proposes the following four changes to the Palisades TSs:
    (1) In section 5.2, ``Organization,'' Palisades TS Section 5.5.2e 
would be revised by deleting the specific detail of working hour 
limitations (i.e., administrative procedures are used to control 
working hours).
    (2) Also in Section 5.2, TS Section 5.5.2g would be revised by 
deleting the title for the ``Shift Technical Advisor'' position and by 
clarifying the requirements for that position.
    (3) In TS Section 5.5.4, ``Radioactive Effluent Controls Program,'' 
sections 5.5.4b, 5.5.4e, and 5.5.4h would be revised to be consistent 
with 10 CFR part 20.
    (4) TS Section 5.7, ``High Radiation Area,'' would be revised to be 
consistent with 10 CFR Part 20.1601(c) (i.e., the existing TS would be 
completely replaced by Insert F from TSTF 258).
    Additional changes requested in the licensee's application dated 
December 7, 2000, are based upon other TSTFs and are addressed by 
separate Federal Register notices.
    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 changes would not] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    All four proposed changes deal exclusively with Administrative 
Controls. The changes only affect the details of controls placed on 
the plant staff and their working conditions. The proposed controls 
are consistent with the requirements approved for STS. None of the 
controls involved are assumed to be associated with any initiator 
of, or any mitigating equipment or mitigation actions for any 
accident previously evaluated. Therefore, the proposed changes will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    B. [The proposed changes would not] create the possibility of a 
new or different kind of accident from any previously evaluated.
    All four proposed changes deal exclusively with Administrative 
Controls. The changes only affect the details of controls placed on 
the plant staff and their working conditions. The proposed controls 
are consistent with the requirements approved for STS. The proposed 
changes would not result in any physical alterations to the plant 
configuration, no new equipment is added, no equipment interfaces 
are modified, no changes to any equipment's function or the method 
of operating the equipment are being made. As the proposed changes 
would not change the design, configuration or operation of the 
plant, no new or different kinds of accident modes are created. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    C. [The proposed changes would not] involve a significant 
reduction in a margin of safety.
    All four proposed changes deal exclusively with Administrative 
Controls. The changes only affect the details of controls placed on 
the plant staff and their working conditions. The proposed controls 
are consistent with the requirements approved for STS. None of the 
controls involved are assumed to be associated with any initiator 
of, or any mitigating equipment or mitigation actions for any 
accident previously evaluated. 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: December 7, 2000.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) in accordance with changes to 
the ``Standard Technical Specifications, Combustion Engineering 
Plants,'' NUREG 1432, Revision 1 (STS), made by the Nuclear Energy 
Institute Technical Specifications Task Force (TSTF) change number 287, 
``Allowances For Breach Of The Control Room Envelope,'' Revision 5. 
Specifically, a note would be added modifying TS section 3.7.10, 
``Control Room Ventilation (CRV) Filtration,'' to allow the control 
room boundary to be opened intermittently under administrative control, 
and a new condition (Condition B) would be added to the Action table 
for TS section 3.7.10 to allow 24 hours to restore an inoperable 
control room boundary. A required Action (B.1) would also be added 
requiring certain preplanned actions to be initiated immediately upon 
discovery that the containment envelope is inoperable. The subsequent 
conditions and required actions would be renumbered accordingly and 
supporting editorial changes would be made to the descriptions for 
Conditions B and E (to be renumbered as Conditions C and F). The 
licensee also forwarded related changes to the Bases for TS section 
3.7.10.
    Additionally, a correction would be made to the Action table for TS 
Section 3.7.10 by restoring Required Action D.2 (to be renumbered to 
E.2), which was inadvertently omitted during the prior issuance of the 
Palisades Improved TSs by Amendment No. 189.
    Additional changes requested in the licensee's application dated 
December 7, 2000, are based upon other TSTFs and are addressed by 
separate Federal Register notices.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 7679]]


    A. [The proposed changes would not] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes deal exclusively with allowances to 
temporarily deviate from the [Limiting Condition for Operation] LCO 
3.7.10 requirement (established by [Surveillance Requirement] SR 
3.7.10.4) for the control room boundary to be sufficiently air tight 
to maintain 0.125 inches of water differential when the ventilation 
system is in the emergency mode of operation. The proposed controls 
are consistent with the requirements approved for STS. None of the 
controls involved are assumed to be associated with any assumed 
initiator of any accident previously evaluated.
    The proposed changes do allow temporary (up to 24 hours) 
relaxation of controls put in place to protect the operators from 
accidental releases of particulate radioactive materials. The 
utilization of this temporary allowance is expected to be 
infrequent, and the controls required when this allowance is 
utilized maintain the intended radiological protection for the 
operators in the control room areas. Since the protection of the 
operators in the control room areas will be provided by alternate 
means during the exercising of these allowances, there will be no 
effect on their perceived abilities to mitigate the consequences of 
an accident.
    Therefore, operation of the facility in accordance with the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    B. [The proposed changes would not] create the possibility of a 
new or different kind of accident from any previously evaluated.
    The proposed changes deal exclusively with an allowance to 
temporarily provide radiological protection within the control room 
boundary by alternative means. The proposed controls are consistent 
with the requirements approved for STS. The proposed changes would 
not result in any physical alterations to the operating plant 
systems, no new equipment is added, no equipment interfaces are 
modified, no changes to any equipment's function or the method of 
operating the power generation or accident mitigating equipment are 
being made. As the proposed changes would not change the design, 
configuration or operation of the plant, no new or different kinds 
of accident modes are created. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    C. [The proposed changes would not] involve a significant 
reduction in a margin of safety.
    The proposed changes deal exclusively with an allowance to 
temporarily provide radiological protection within the control room 
boundary by alternative means. The proposed controls are consistent 
with the requirements approved for STS. None of the controls 
involved are assumed to be associated with any initiator of, or any 
mitigating equipment or mitigation actions for any accident 
previously evaluated. 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: December 8, 2000.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) (and, as 
applicable, other elements of the licensing bases) to maintain a post 
accident sampling system (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271), on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated December 8, 2000.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase

[[Page 7680]]

in the consequences of any accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: October 24, 2000.
    Description of amendment request: Entergy Operations, Inc. is 
proposing that the Grand Gulf Nuclear Station (GGNS) Operating License 
be amended to revise the GGNS Technical Specifications (TSs), which 
govern the lube oil inventories for the Division I, II, and III 
Emergency Diesel Generators (EDGs). The change would increase the lube 
oil inventories specified in TS 3.8.3 to ensure continued operation of 
the EDGs under post-accident conditions, and provide additional margin 
in lube oil consumption calculations. The TS change would account for 
potential increases in EDG lube oil consumption rates which exceed the 
nominal consumption rates originally used to determine EDG lube oil 
requirements to support seven days of EDG operation at rated load 
conditions.
    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?
    The purpose of the emergency diesel generators is to mitigate 
the consequences of analyzed accidents. Emergency Diesel Engine 
inoperability or loss of capability has no effect on the probability 
of any analyzed accident. The reason for this change is to provide 
added assurance that the engines perform per the design requirements 
and therefore the consequences of an accident previously evaluated 
are not increased.
    The purpose of the requested change is to regain margin in the 
lube oil consumption calculations, such that, if increases in 
consumption should occur in the future, Technical Specifications 
requirements will still ensure operability of the Diesel Generators. 
Design Engineering has basically taken the vendor's specified 
consumption rate and doubled that value to ensure that the newly 
calculated inventory limit will bound any potential consumption rate 
increases.
    Current calculations using as found consumption rates have shown 
that the limiting sump volume is on Division III engines and that 
there is minimal margin left between the actual volume and the 
calculated volume needed. Therefore, there is a need for an external 
dedicated storage skid, which is the only physical change to the 
plant necessary to support this change request. The current 
licensing basis recognizes that make-up oil may be required at some 
point during a design basis event. The current Bases for Technical 
Specification 3.8.3 LCO provides this recognition.
    Given the stated purpose and no need for changes to installed 
plant structures, (other than addition of a new Div[ision] III lube 
oil storage skid) systems, or components there will be no 
significant changes to the operation of the facility. Therefore, 
this change does 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 kind of 
accident from any accident previously evaluated?
    The purpose of the emergency diesel generators is to mitigate 
the consequences of analyzed accidents; the engines are not accident 
initiating. Emergency Diesel Engine inoperability or loss of 
capability cannot create the possibility of a new or different kind 
of accident from any accident previously evaluated. The reason for 
this change is to provide added assurance that the engines perform 
per the design requirements.
    The Diesel Engine Lubricating System (DELS) design and operation 
is unaffected by his change. Recognizing the need for having a make-
up inventory and staging a volume readily accessible to the operator 
will enhance the operator's ability to maintain DG [Diesel 
Generator] operable. Design Engineering has performed appropriate 
fire hazards reviews and seismic II/l reviews to assure compliance 
with current design requirements.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The current licensing basis requires that the DELS provide seven 
days of Diesel Generator operation under specified load conditions. 
This basis was substantiated via calculation using vendor supplied 
consumption rates of 1.21 (Div[ision] I and II) and 0.6 (Div[ision] 
Ill) gallons per hour. The current basis recognizes that make-up oil 
may be required at some point during a design basis event. To ensure 
this basis is valid for future operations, Design Engineering has 
recalculated the required inventories based on a more conservative 
consumption rate. This change will ensure that sufficient lube oil 
is readily available to support the extended run times under post 
accident conditions. Therefore, this change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 21, 2000.
    Description of amendment request: The amendment request proposes to 
delete the Steam/Feedwater Flow Mismatch coincident with Low Steam

[[Page 7681]]

Generator (SG) Water Level reactor trip from the technical 
specifications. The Steam/Feedwater Flow Mismatch coincident with Low 
SG Water Level reactor trip was included in the Unit 1 design in order 
to meet regulatory requirements regarding potentially adverse control 
and protection system interactions. The amendment request proposes to 
take credit for the SG Level Median Selector Switch (MSS) installed in 
1997 to meet these requirements. The MSS eliminates the potential for 
an adverse control and protection system interaction and, therefore, 
eliminates the design requirement for the Steam/Feedwater Flow Mismatch 
and Low SG Level reactor trip. Appropriate changes to the Bases are 
also included in the amendment request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The initiating conditions and assumptions for accidents 
described in the Updated Final Safety Analys[i]s Report remain as 
previously analyzed. The proposed change does not introduce a new 
accident initiator nor does it introduce changes to any existing 
accident initiators or scenarios described in the Updated Final 
Safety Analys[i]s Report. The Steam/Feedwater Flow Mismatch and Low 
Steam Generator Water Level reactor trip is not credited for 
accident mitigation in any accident analyses described in the 
Updated Final Safety Analys[i]s Report. The Steam/Feedwater Flow 
Mismatch and Low Steam Generator Water Level trip was designed to 
meet the control and protection systems interaction criteria of the 
Institute of Electric and Electronic Engineers Standard 279. The 
Median Selector Switch prevents adverse control and protection 
system interaction such that it replaces the need for the Steam/
Feedwater Flow Mismatch and Low Steam Generator Water Level reactor 
trip and satisfies the Institute of Electric and Electronic 
Engineers Standard 279 requirements. As such, the affected control 
and protection systems will continue to perform their required 
functions without adverse interaction and the capability to shut 
down the reactor when required on Low-Low Steam Generator water 
level to mitigate an accident previously evaluated is unaffected.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The substitution of the Median Selector Switch for the Steam/
Feedwater Flow Mismatch and Low Steam Generator Water Level trip 
will not introduce any new failure modes to the required protection 
functions. The Median Selector Switch only interacts with the 
feedwater control system and the Steam Generator Water Level Low-Low 
protection function is not affected by this change. Isolation 
devices in the Median Selector Switch circuitry ensure that the 
Steam Generator Water Level Low-Low protection function is not 
affected. The Median Selector Switch is designed to reduce the 
frequency of system failures through utilization of highly reliable 
components in a design that relies on a minimum of additional 
equipment. Components utilized in the Median Selector Switch are of 
a quality consistent with low failure rates and minimum maintenance 
requirements, and conform to protection system requirements. 
Furthermore, the design provides the capability for complete unit 
testing that provides unambiguous determination of credible system 
failures. It is through these features that the overall design of 
the Median Selector Switch minimizes the occurrence of undetected 
failures that may exist between test intervals. Additionally, the 
reliability of the Median Selector Switch has been shown by Unit 2 
operating experience to be acceptable.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety depends on the maintenance of specific 
operating parameters and systems within design requirements and 
safety analysis assumptions.
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not alter the functional 
capabilities assumed in a safety analysis for any system, structure, 
or component important to the mitigation and control of design bases 
accident conditions within the facility. Nor does this amendment 
revise any parameters or operating restrictions that are assumptions 
of a design basis accident. In addition, the proposed amendment does 
not affect the ability of safety systems to ensure that the facility 
can be placed and maintained in a shutdown condition for extended 
periods of time.
    The ability of the Steam Generator Water Level Low-Low reactor 
trip function credited in the safety analysis to protect against a 
sudden loss of heat sink event is not affected by the proposed 
change. Since the Steam Generator Low-Low Level trip provides 
complete protection for all accident transients that result in low 
steam generator level, eliminating the Steam/Feedwater Flow Mismatch 
and Low Steam Generator Water Level trip will not change any safety 
analysis conclusion for any analyzed accident described in the 
Updated Final Safety Analys[i]s Report.
    The Median Selector Switch prevents adverse control and 
protection system interaction such that it replaces the need for the 
Steam/Feedwater Flow Mismatch and low Steam Generator Water Level 
reactor trip and satisfies the Institute of Electric and Electronic 
Engineers Standard 279 requirements. The proposed change will 
enhance safe operation since the Steam/Feedwater Flow Mismatch and 
Low Steam Generator Water Level trip function removal decreases the 
challenges to the plant safety systems, decreases the plant 
surveillance/maintenance activity, and reduces the plant complexity; 
all resulting in a reduction in the potential for unnecessary plant 
transients.
    The technical specifications continue to assure the applicable 
operating parameters and systems are maintained within the design 
requirements and safety analysis assumptions. Therefore, the 
elimination of this trip function will not result in a significant 
reduction in the margin of safety as defined in the Updated Final 
Safety Analys[i]s Report or technical specifications.

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

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: January 2, 2001.
    Description of amendment requests: The proposed amendment would 
revise Technical Specifications (TS) 3/4.6.2.2.a for the Unit 1 spray 
additive tank to require a contained volume between 4000 and 4600 
gallons of between 30 and 34 percent by weight sodium hydroxide (NaOH) 
solution. In addition, the proposed amendment would make four types of 
format changes to the revised Unit 1 page:
    1. Reformat the header to include numbered first and second tier TS 
section titles and a full-width single line to separate the header 
section titles from the page text.
    2. Reformat the footer to include ``COOK NUCLEAR PLANT--UNIT1'' on 
the left side of the page, ``Page (page number)'' center page, 
``AMENDMENT (past amendment numbers, with strikethrough, and ending 
with the current amendment number)'' on the right side, and a full-
width single line to separate the footer from the page text.
    3. Delete the double lines under ``LIMITING CONDITION FOR 
OPERATION'' and ``SURVEILLANCE REQUIREMENTS.''
    4. Fully justify the text and change the font.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 7682]]

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

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    Adding a maximum limit for the allowed contained volume and 
[sodium hydroxide] NaOH concentration for the spray additive tank 
does not increase the probability of occurrence of any accident. The 
spray additive system cannot initiate any previously analyzed 
accident. The proposed changes ensure that the spray additive system 
and the associated containment spray system can perform the accident 
mitigation functions required during a [loss-of-coolant accident] 
LOCA or [main steam line break] MSLB event. This action does not 
affect the initiating frequency of a LOCA or MSLB event. Therefore, 
the proposed changes do not increase the probability of an accident 
previously evaluated.
    The accidents previously evaluated in Chapter 14 of the Updated 
Final Safety Analysis Report that are possibly affected by operation 
of the spray additive system are a loss-of-coolant accident (LOCA) 
and a main steam line break (MSLB). These postulated accidents are 
expected to result in a containment spray signal, which then results 
in the automatic starting of the containment spray pumps and the 
opening of the valves associated with the spray additive system. The 
spray additive system adds NaOH to the containment spray water being 
supplied from the refueling water storage tank (RWST) to adjust the 
pH of the containment spray and containment recirculation sump 
solutions.
    Following a LOCA, the containment spray water becomes mixed in 
the containment recirculation sump with ice melt from the ice 
condenser, reactor coolant from the reactor coolant system (RCS), 
water being injected to the RCS from the safety injection 
accumulators, and water being injected to the RCS from the RWST by 
the emergency core cooling system. Following a MSLB, the containment 
spray water becomes mixed in the containment recirculation sump with 
ice melt from the ice condenser and the secondary coolant released 
from the ruptured steam line.
    The existing minimum and proposed maximum limits for the 
contained volume and NaOH concentration for the spray additive tank 
ensure a pH value of between 7.6 and 9.5 for the solution 
recirculated within containment after a LOCA. This pH band minimizes 
the evolution of iodine from the containment recirculation sump, and 
minimizes the effect of chloride and caustic stress corrosion on 
mechanical systems and components. An increase in pH value to at 
least 7.0 in the containment recirculation sump during the 
recirculation phase following a LOCA is consistent with the iodine 
retention assumptions of the accident analyses. Therefore, the 
consequences of a LOCA remain unchanged by the proposed changes. For 
a MSLB, there is no increase in consequences since the containment 
spray system and containment recirculation sump are not credited for 
removal and retention of fission products from the containment 
atmosphere.
    The analyses for determining hydrogen generation following a 
large break LOCA assume a specific pH time-dependent profile for the 
containment spray and containment recirculation sump solutions. The 
existing minimum and proposed maximum limits for the contained 
volume and NaOH concentration for the spray additive tank do not 
result in an increase in the previously predicted hydrogen 
generation rates. Therefore, the current hydrogen generation 
analyses remain bounding.
    For both LOCA and MSLB events, the existing minimum and proposed 
maximum limits for the contained volume and NaOH concentration for 
the spray additive tank ensure that the pH of the containment spray 
solution is within the bounds used in evaluations for environmental 
qualification of required equipment.
    Therefore, the proposed changes cannot increase the probability 
of occurrence or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Adding a maximum allowed contained volume and NaOH concentration 
for the spray additive tank does not create the possibility of an 
accident of a new or different type than any previously evaluated. 
The proposed changes ensure that the spray additive system, and the 
associated containment spray system, can perform the required 
accident mitigation functions during a LOCA or MSLB event. There are 
no other types of accidents that can be postulated that would 
require the use of the spray additive system or the associated 
containment spray system for mitigation. The proposed changes do not 
introduce any new association between the spray additive system and 
any radioactive system, including the RCS. Therefore, emergency 
operation of the spray additive system, or postulated failures of 
the spray additive system, cannot initiate any type of accident.
    Therefore, the proposed changes do not increase the possibility 
of a new or different kind of accident than previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed limits on maximum allowed contained volume and NaOH 
concentration for the spray additive tank ensure that the original 
margin of safety is maintained by ensuring acceptable pH control 
following a LOCA or MSLB event. Therefore, the proposed changes 
ensure that the margin of safety is maintained by limiting the 
maximum pH of the containment spray and containment recirculation 
sump solutions following a LOCA or MSLB event.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

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

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: October 24, 2000.
    Description of amendment requests: The proposed amendments would 
approve an unreviewed safety question allowing the use of new 
methodology to calculate the transient response to steam generator tube 
ruptures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed change, to adopt a new analytical method to 
evaluate the effects of an [Steam Generator Tube Rupture] SGTR, does 
not affect any accident initiators or precursors. As such, the 
proposed change does not increase the probability of an accident. 
The proposed change also does not affect the ability of operators to 
mitigate the consequences of an accident. The proposed change does 
not impact the design of the affected plant systems such that 
previously analyzed systems, structures, and components (SSCs) would 
now be more likely to fail. The changes will not modify plant 
systems to reduce their design capability during normal operating 
and accident conditions. The use of the WCAP-10698-P-A methodology 
to more accurately calculate the flow from the reactor coolant 
system (RCS) to the SG secondary side following a postulated SGTR 
does not affect the probability of any analyzed events. The use of 
the WCAP-10698-P-A methodology does not affect SGTR initiators or 
precursors. Therefore, incorporating the new methodology does not 
affect equipment malfunction probability, nor does it affect or 
create new accident initiators or precursors. Thus, there will be no 
reduction in the capability of those SSCs in limiting the 
consequences of previously evaluated accidents.
    Additionally, the present methodology for calculating the 
radiological consequences of a postulated SGTR is conservative when 
compared with results from the new methodology. As such, the 
existing licensing basis radiological consequence calculations will 
be retained. Thus, no additional radiological source terms are 
generated, and the consequences of an accident previously

[[Page 7683]]

evaluated in the [updated final safety analysis report] UFSAR will 
not be increased. The use of this WCAP methodology and associated 
computer code for break flow modeling more accurately calculates the 
plant response to an SGTR event. The improved accuracy of the new 
methodology provides valuable information related to the analysis of 
operator actions and the associated timing. Such accurate transient 
response information enables enhancements to be made to the 
emergency operating procedures (EOPs).
    Therefore, the proposed changes cannot increase the consequences 
or probability of occurrence of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not impact the design of affected plant 
systems, involve a physical alteration to the systems, or change to 
the way in which systems are currently operated, such that 
previously unanalyzed SGTRs would now occur. The change to 
incorporate the WCAP-10698-P-A methodology does not introduce any 
new malfunctions; it calculates more accurately the flow from the 
RCS to the SG secondary side following a postulated SGTR to 
determine the time available for operator actions to prevent 
overfilling the affected SG.
    Thus, use of the WCAP-10698-P-A methodology does not affect or 
create new accident initiators or precursors or create the 
possibility of a new or different kind of accident.
    Therefore, the proposed changes do not increase the possibility 
of a new or different kind of accident than previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The approval of the license amendment will not result in any 
modifications to affected plant systems that would reduce their 
design capabilities during normal operating and accident conditions. 
By using the WCAP-10698-P-A methodology, a more accurate SGTR 
response is calculated. The improved understanding of the transient 
response enables enhancements to the EOPs, which provide further 
assurance that SSCs required for accident mitigation are protected.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    In summary, based upon the above evaluation, I&M has concluded 
that these changes involve no significant hazards.

    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.

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

    Date of amendment request: December 18, 2000.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated December 18, 2000.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result of 
the TMI-2 accident. The specific intent of the PASS was to provide a 
system that has the capability to obtain and analyze samples of plant 
fluids containing potentially high levels of radioactivity, without 
exceeding plant personnel radiation exposure limits. Analytical results 
of these samples would be used largely for verification purposes in 
aiding the plant staff in assessing the extent of core damage and 
subsequent offsite radiological dose projections. The system was not 
intended to and does not serve a function for preventing accidents and 
its elimination would not affect the probability of accidents 
previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual benefit 
to post accident mitigation. Past experience has indicated that there 
exists in-plant instrumentation and methodologies available in lieu of 
a PASS for collecting and assimilating information needed to assess 
core damage following an accident. Furthermore, the implementation of 
Severe Accident Management Guidance (SAMG) emphasizes accident 
management strategies based on in-plant instruments. These strategies 
provide guidance to the plant staff for mitigation and recovery from a 
severe accident. Based on current severe accident management strategies 
and guidelines, it is determined that the PASS provides little benefit 
to the plant staff in coping with an accident.
    The regulatory requirements for the PASS can be eliminated without 
degrading the plant emergency response. The emergency response, in this 
sense, refers to the methodologies used in ascertaining the condition 
of the reactor core, mitigating the consequences of an accident, 
assessing and projecting offsite releases of radioactivity, and 
establishing protective action recommendations to be communicated to 
offsite authorities. The elimination of the PASS will not prevent an 
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency 
plan (EP), the emergency operating procedures (EOP), and site survey 
monitoring that support modification of emergency plan protective 
action recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical

[[Page 7684]]

Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any accident 
previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.
    The elimination of PASS related requirements will not result in any 
failure mode not previously analyzed. The PASS was intended to allow 
for verification of the extent of reactor core damage and also to 
provide an input to offsite dose projection calculations. The PASS is 
not considered an accident precursor, nor does its existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within the 
containment building.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a neutral 
impact to the margin of safety. Methodologies that are not reliant on 
PASS are designed to provide rapid assessment of current reactor core 
conditions and the direction of degradation while effectively 
responding to the event in order to mitigate the consequences of the 
accident. The use of a PASS is redundant and does not provide quick 
recognition of core events or rapid response to events in progress. The 
intent of the requirements established as a result of the TMI-2 
accident can be adequately met without reliance on a PASS.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    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: July 31, 2000.
    Description of amendment request: The proposed license amendment 
will change the method used to determine the Fuel Centerline Melt 
Linear Heat Rate Limit (FCMLHRL). The proposed change represents a 
departure from the use of the fixed value of 21 kilowatts per foot for 
the FCMLHRL, which is being used in the current operating cycle, to a 
value that will be calculated on a cycle-by-cycle basis using the 
Siemens Power Corporation (SPC) U.S. Nuclear Regulatory Commission 
(NRC) approved methodology. Northeast Nuclear Energy Company (the 
licensee) has evaluated this proposed method of calculating FCMLHRL 
utilizing the criteria of 10 CFR 50.59. The licensee has determined 
that this change involves an unreviewed safety question (USQ). The 
licensee is requesting approval of the USQ.
    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.
    This license amendment request deals with changes in the 
Millstone Unit No. 2 Final Safety Analysis Report (FSAR) due to 
changing the method used to determine the FCMLHRL. The proposed 
change represents a departure from the use of the fixed value of 21 
kW/ft for the FCMLHRL, which is being used in the current cycle, to 
a value that will be calculated on a cycle by cycle basis using the 
SPC approved methodology. This methodology was reviewed and approved 
by the Nuclear Regulatory Commission (NRC) and is documented in 
Siemens Power Corporation (SPC) report XN-NF-82-06(P)(A).[ ] The 
value of the FCMLHRL is verified for each reload, but does not 
typically change significantly between cycles. This limit is 
determined for a standard fuel rod. The current enrichment cutbacks 
in the gadolinia bearing rods limit their relative power such that 
the maximum FCMLHRL for a gadolinia bearing fuel rod will be 
sufficiently below the standard fuel rods to prevent centerline 
melt. In future applications of this methodology, the peak Linear 
Heat Rates (LHR) calculated from transient analyses will be compared 
to the FCMLHRL for the cycle. The Local Power Density (LPD) Limiting 
Safety System Settings (LSSS) verification analysis for future 
applications will use the cycle dependent FCMLHRL. Therefore, It can 
be concluded that these FSAR changes are safe and that the cycle 
specific calculated FCMLHRL has no impact on plant equipment 
operation. Further more, the change in the method of determining the 
FCMLHRL only impacts the analytical determination of failed fuel and 
has no direct impact on the accident scenario. Accordingly, this 
change cannot affect the likelihood of these events. Therefore, the 
proposed changes will not increase the probability of occurrence of 
accidents previously evaluated.
    The change in the method of determining the FCMLHRL will 
continue to conservatively estimate fuel failures. Since the 
proposed FSAR changes will have no impact on the analysis of the 
events, they cannot affect the likelihood or consequences of these 
events. Therefore, the proposed FSAR changes will not increase the 
consequences of accidents previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed FSAR 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. The FSAR changes do not 
introduce any new failure modes. Therefore, the changes will not 
increase the probability of a new or different kind of accident from 
any accidents previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The purpose of the proposed changes is to document a change in 
the method used to determine FCMLHRL in the Millstone Unit No. 2 
FSAR. The change in methodology may result in a FCMLHRL that is 
higher than the previous limit of 21 kW/ft. Therefore, the proposed 
changes may lead to a reduction of the margin of safety. However, 
the proposed changes are safe because SPC has justified, using NRC 
generically approved methodology, that with a higher value of the 
FCMLHRL the fuel will not experience centerline melt. In other 
words, a higher FCMLHRL may allow a higher fuel temperature but will 
continue to protect fuel against centerline melt. Therefore, it can 
be concluded that the FSAR changes are safe and do not significantly 
reduce the margin of safety.

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

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: December 13, 2000.
    Description of amendment request: The proposed amendment would

[[Page 7685]]

change License Condition 2.C.4 to conform to NRC Generic Letter (GL) 
86-10, ``Implementation of Fire Protection Requirements.'' The proposed 
amendment would also relocate the Fire Protection Program (FPP) 
elements from the Technical Specifications (TSs) to the licensee-
controlled FPP, in accordance with GL 86-10 and GL 88-12, ``Removal of 
Fire Protection Requirements from 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 proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The requested changes are administrative in nature in that they 
move fire protection requirements from the TS to the FPP and 
associated implementing procedures following the guidance of NRC 
Generic Letter (GL) 86-10 and GL 88-12. The requested changes will 
not revise the requirements for fire protection equipment 
operability, testing or inspections.
    The proposed changes do not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident, nor do they affect 
any assumptions or conditions in any of the accident analyses. Since 
the accident analyses remain bounding, their radiological 
consequences are not adversely affected.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The requested changes are administrative in nature in that they 
move fire protection requirements from the TS to the FPP and 
associated implementing procedures following the guidance of GL 86-
10 and 88-12. The requested changes will not revise the requirements 
for fire protection equipment operability, testing or inspections.
    The proposed changes do not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident, nor do they affect 
any assumptions or conditions in any of the accident analyses. 
Accordingly, no new failure modes have been defined for any plant 
system or component important to safety nor has any new limiting 
single failure been identified as a result of the proposed changes.
    Therefore the possibility of a new or different kind of accident 
from any accident previously evaluated is not created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The requested changes are administrative in nature in that they 
move fire protection requirements from the TSs to the FPP and 
associated implementing procedures following the guidance of GL 86-
10 and 88-12. The requested changes will not revise the requirements 
for fire protection equipment operability, testing or inspections. 
Future changes to the program will be reviewed in accordance with 
the fire protection license condition to ensure that the ability to 
achieve and maintain safe shutdown in the event of a fire are [sic] 
not adversely affected.
    Therefore, a significant reduction in the margin of safety is 
not involved.

    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 E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: December 13, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.8/4.8 to clarify the air ejector 
offgas activity sample point and operability requirements.
    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 proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes clarify and more completely specify actions and 
requirements with respect to main condenser offgas activity. 
Compliance with applicable regulatory requirements will continue to 
be maintained. The proposed changes do not alter the conditions or 
assumptions in any of the previous accident analyses. Since the 
previous accident analyses remain bounding, the radiological 
consequences previously evaluated are not adversely affected by the 
proposed changes.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes do not involve any change to the method of 
operation of any plant equipment. Accordingly, no new failure modes 
have been defined for any plant system or component important to 
safety nor has any new limiting single failure been identified as a 
result of the proposed changes. Also, there will be no change in 
types or increase in the amounts of any effluents released offsite.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes clarify and more completely 
specify actions and requirements with respect to main condenser 
offgas activity. No changes in radioactivity release limits or dose 
limits are proposed. The changes in actions to be taken if a limit 
is not met provide an adequate means of ensuring that the health and 
safety of the public are protected and that potential dose to the 
public is below regulatory limits. The proposed changes do not 
involve any actual change in the methodology used in the control of 
radioactive effluents. The proposed changes also comply with the 
guidance contained in the STS [standard technical specifications].
    Therefore, a significant reduction in the margin of safety is 
not involved.

    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 E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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 amendment requests: December 6, 2000.
    Description of amendment requests: The proposed license amendments 
would revise Section 5.0, ``Administrative Controls,'' of the Diablo 
Canyon Power Plant, Unit Nos. 1 and 2 Technical Specifications (TS) to 
change the following management titles.
    (1) TS 5.1.1 would be revised to replace the titles ``Vice 
President, Diablo Canyon Operations and Plant Manager,'' and ``Plant 
Manager,'' with the generic title ``plant manager.''

[[Page 7686]]

    (2) TS 5.2.1.a, last sentence, would be revised to state: ``These 
requirements, including the plant-specific titles of those personnel 
fulfilling the responsibilities of the positions delineated in these 
Technical Specifications, shall be documented in the FSAR [Final Safety 
Analysis Report] Update.''
    (3) TS 5.2.1.b, would be revised to replace the title ``Plant 
Manager,'' with the generic title ``plant manager.''
    (4) TS 5.2.1.c. would be revised to replace the title ``Senior Vice 
President and General Manager--Nuclear Power Generation,'' with the 
generic title ``specified corporate officer.''
    (5) TS 5.2.2.d would be revised to replace the title ``Plant 
Manager,'' with the generic title ``plant manager.''
    (6) TS 5.2.2.e. would be revised to replace the title ``Operations 
Director'' with the generic title ``operations manager.''
    (7) TS 5.3.1 would be revised to replace the titles ``Radiation 
Protection Director'' and ``Operations Director'' with the generic 
titles ``radiation protection manager'' and ``operations manager,'' 
respectively.
    (8) TS 5.5.1.b (second paragraph ``b'') would be revised to replace 
the title ``Plant Manager,'' with the generic title ``plant manager.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This License Amendment Request (LAR) proposes to revise 
Technical Specification (TS) 5.0, ``Administrative Controls,'' to 
replace specific management titles with lower case generic titles 
consistent with Industry/Technical Specification Task force (TSTF) 
Standard Technical Specification Change Traveler TSTF-65, Revision 
1, approved by the NRC on November 10, 1994.
    The proposed changes revise TS 5.0 to change management titles 
from (a) ``Vice President, Diablo Canyon Operations and Plant 
Manager'' to ``plant manager,'' (b) ``Senior Vice President and 
General Manager--Nuclear Power Generation'' to ``specified corporate 
officer,'' (c) ``Radiation Protection Director'' to ``radiation 
protection manager,'' and (d) ``Operations Director'' to 
``operations manager.''
    The proposed changes do not eliminate any of the qualifications, 
responsibilities or requirements for these positions. Each member of 
the plant staff assigned to these positions shall continue to meet 
or exceed the minimum qualifications of ANSI/ANS 3.1-1978, 
Regulatory Guide 1.8, Revision 2, April 1987 (radiation protection 
manager), or TS 5.2.2.e (operations manager) as required by TS 
5.3.1.
    The proposed change to replace the title ``Vice President, 
Diablo Canyon Operations and Plant Manager'' with the generic title 
``plant manager'' reflects PG&E's plan to split the responsibilities 
of the Vice President, Diablo Canyon Operations and Plant Manager, 
into two positions: (1) Vice President, Diablo Canyon Operations, 
and (2) Station Director. The Station Director will report to the 
Vice President, Diablo Canyon Operations. The Station Director will 
fulfill the responsibilities of the ``Plant Manager'' as described 
currently in TS and Final Safety Analysis Report (FSAR) Update and 
will be responsible for overall safe operation of the plant and will 
have control over those onsite activities necessary for safe 
operation and maintenance of the plant. This change results in no 
change to the responsibilities or qualification requirements for 
this position as specified in the TS.
    The remaining changes are administrative changes only that 
result in no changes in the responsibilities for the positions.
    None of the proposed changes have an impact on plant equipment, 
or on how plant equipment is operated or maintained, and therefore 
they have no impact on plant accidents.
    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 revise TS 5.0 to change management titles 
from (a) ``Vice President, Diablo Canyon Operations and Plant 
Manager'' to ``plant manager,'' (b) ``Senior Vice President and 
General Manager--Nuclear Power Generation'' to ``specified corporate 
officer,'' (c) ``Radiation Protection Director'' to ``radiation 
protection manager,'' and (d) ``Operations Director'' to 
``operations manager.''
    The proposed changes do not eliminate any of the qualifications, 
responsibilities or requirements for these positions.
    None of the proposed changes have an impact on plant equipment, 
or on how plant equipment is operated or maintained, and therefore 
they have no impact on initiation of new or different plant 
accidents.
    Therefore, the proposed changes do 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 revise TS 5.0 to change management titles 
from (a) ``Vice President, Diablo Canyon Operations and Plant 
Manager'' to ``plant manager,'' (b) ``Senior Vice President and 
General Manager--Nuclear Power Generation'' to ``specified corporate 
officer,'' (c) ``Radiation Protection Director'' to ``radiation 
protection manager,'' and (d) ``Operations Director'' to 
``operations manager.''
    The proposed changes do not eliminate any of the qualifications, 
responsibilities or requirements for these positions.
    None of the proposed changes have an impact on plant equipment, 
or on how plant equipment is operated or maintained, and therefore 
they have no impact on margin of safety.
    Therefore, the proposed changes do not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: December 7, 2000.
    Description of amendment request: The licensee proposes to revise 
Technical Specification 3.5.A.1 by adding a note regarding operability 
of the Low Pressure Coolant Injection system (LPCI) under certain 
restrictive conditions. The subject change would provide a 
clarification of system operability that would result in additional 
flexibility in operations during hot shutdown conditions.
    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 LPCI system is not assumed to be the initiator of any 
previously analyzed event. Its function is in mitigating and thereby 
limiting consequences of analyzed events. With this proposed change 
LPCI is still capable of being manually realigned, if needed, to 
mitigate the consequences of accidents. The allowance provided by 
this change is only applicable for the reactor in a shutdown 
condition with reactor pressure less than the RHR [residual heat 
removal] shutdown cooling permissive setpoint.
    Thus, the reactor heat load is much less than assumed for design 
basis loss of coolant accidents occurring at full power. 
Furthermore, other emergency core cooling systems are still required 
to be operable.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 7687]]

    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not involve any physical alteration of 
the plant or introduce new modes of operation. There is no change in 
plant operation that involves failure modes other than those 
previously evaluated.
    The methods governing plant operation and testing remain 
consistent with current safety analysis assumptions. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed change has no impact on any safety analysis 
assumption. The clarifying Note being added to Technical 
Specification 3.5.A.1 allows the decay heat removal function to be 
available without immediate shutdown requirements for inoperable 
LPCI subsystems being imposed. This is recognition that the amount 
of time to realign the RHR system from the decay heat removal 
function has no significant impact on the margin of safety 
associated with establishing LPCI injection, because the heat loads 
under these conditions are far less than assumed in the safety 
analysis.
    Placing the reactor in SDC [Shutdown Cooling] during hot 
shutdown is a normal and preferred method for removing sensible heat 
from the reactor. In addition, the change does not alter the 
availability of other safety systems and the ability to meet their 
safety functions. The additional flexibility, to allow LPCI 
subsystems to be considered operable during SDC below the RHR 
shutdown cooling permissive pressure and without entering a shutdown 
LCO [limiting condition for operation] will not significantly reduce 
margins of safety since the reactor is in hot shutdown with all 
control rods inserted, reactor pressure is less than the RHR 
shutdown cooling permissive pressure, and other ECCS [Emergency Core 
Cooling Systems] systems should be capable of providing the required 
cooling, thereby allowing operation of RHR SDC when necessary. Thus, 
the margins of safety for such situations are maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: December 19, 2000.
    Description of amendment request: The proposed change would revise 
the reactor vessel pressure/temperature (P/T) limit curves specified in 
TS 3.6.A.1, ``Reactor Coolant Systems--Pressure and Temperature 
Limitations,'' as graphically represented in Figure 3.6.1, for reactor 
heatup, cooldown, and critical operation, as well as for inservice 
hydrostatic and leak tests.
    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 changes to the calculational methodology for the [pressure/
temperature] P/T limits based upon [American Society of Mechanical 
Engineers] ASME [American Society for Mechanical Engineers Boiler 
and Pressure Vessel Code] Code Cases N-640 and N-588 provide 
adequate margin in the prevention of a brittle-type fracture of the 
reactor pressure vessel (RPV). The Code Cases were developed based 
upon the knowledge gained through years of industry experience. The 
experience gained in the areas of fracture toughness of materials 
and pre-existing undetected defects show that some of the existing 
assumptions used for the calculation of P/T limits are unnecessarily 
conservative and unrealistic. Therefore, providing the allowances of 
the subject Code Cases in developing the P/T limit curves will 
continue to provide adequate protection against nonductile-type 
fractures of the RPV.
    The evaluation for revising the P/T limit curves for 
4.46 x 108MWH(t) (32 effective full power years) was 
performed using the approved methodologies of 10 CFR 50, appendix G. 
The curves generated from these methods ensure the P/T limits will 
not be exceeded during any phase of reactor operation. The proposed 
changes will not affect any other system or equipment designed for 
the prevention or mitigation of previously analyzed events. Thus, 
the probability of occurrence and the consequences of any previously 
analyzed event are not significantly increased as the result of the 
proposed changes.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes to the reactor pressure vessel P/T limits 
do not affect the assumed performance of any system, structure, or 
component previously evaluated. The proposed changes do not 
introduce any new modes of system operation or failure mechanisms. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any 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.
    Industry experience since the inception of the P/T limits in 
1974 confirms that some of the existing methodologies used to 
develop P/T curves is unnecessarily conservative. Accordingly, ASME 
Code Cases N-640 and N-588 take advantage of the acquired knowledge 
by establishing more enhanced methodologies for the development of 
P/T curves. Therefore, operational flexibility can be gained without 
a significant reduction in the margin of safety to RPV brittle 
fracture.
    The revised evaluation of the P/T curves to 
4.46 x 108MWH(t) was performed per the guidelines of 10 
CFR 50, and thus, the margin of safety is not reduced as the result 
of the proposed changes.

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

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: December 14, 2000
    Description of amendment request: The proposed changes will modify 
the Technical Specifications Section 3/4.7.7 ``Control Room Emergency 
Habitability Systems'' Surveillance Requirements 4.7.7.1.d.1 and 
4.7.7.2.a, to revise the differential pressure limit across the control 
room emergency ventilation system filter assembly and increase the 
minimum number of compressed air bottles in the control room bottled 
air pressurization system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    [1.] Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Increasing the minimum required number of air bottles in the 
control

[[Page 7688]]

room bottled air pressurization system in order to maintain system 
capacity does not change the operation of the plant. The control room 
bottled air pressurization system and the emergency ventilation system 
will not be operated differently. No new accident initiators are 
established as a result of the proposed changes. Revising the 
differential pressure acceptance criteria and including [the] demister 
filter along with the HEPA filter and charcoal adsorber will provide 
increased assurance of system readiness. These systems will continue to 
be operable to limit control room dose to within the analysis of 
record. Therefore, the probabaility of occurrence or the consequences 
of an accident previously evaluated is not increased.
    [2.] Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not affect the operation of the plant. The 
control room bottled air pressurization system and control room 
emergency ventilation system will not be operated differently as a 
result of the proposed changes. No new accident or event initiators are 
being created by these changes. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from any 
previously evaluated.
    [3.] Involve a significant reduction in the margin of safety as 
defined in the bases [of] any Technical Specifications.
    The proposed changes reflect conservative changes in the operating 
requirements for the control room bottled air pressurization and 
control room emergency ventilation systems. These changes will further 
ensure the systems will continue to be operable to mitigate the 
consequence of an accident for the control room operators. Therefore, 
the proposed changes do not result in a significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: Richard L. Emch, Jr.

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, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

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

    Date of application for amendment: February 28, 2000, as 
supplemented by letters dated May 12, May 24, June 1, and June 28, 
2000.
    Brief description of amendment: The amendment revised certain 
license conditions to reflect the change in ownership interest from 
PECO to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 228.
    Facility Operating License No. DPR-50: Amendment revised the 
License.
    Date of initial notice in Federal Register: April 10, 2000 (65 FR 
19029).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 2000.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: October 12, 2000.
    Brief description of amendment: The amendment changes the name of 
the facility from WNP-2 to Columbia Generating Station in all 
applicable locations of the Operating License, appendix A Technical 
Specifications, and appendix B Environmental Protection Plan. In 
addition, the proposed action would make editorial changes to TS Figure 
4.1-1, ``Site Area Boundary'' modifying or deleting text associated 
with references to WNP-2.
    Date of issuance: January 8, 2001.
    Effective date: January 8, 2001, and shall be implemented within 30 
days from the date of issuance.
    Amendment No: 169.
    Facility Operating License No. NPF-21: The amendment revised the 
Operating License, appendix A Technical Specifications, and appendix B 
Environmental Protection Plan.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71134).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 23, 1999, as supplemented by 
letters dated February 24 and October 19, 2000.
    Brief description of amendment: The amendment incorporated the use 
of American Society for Testing and Materials (ASTM) D3803-1989, 
``Standard Test Method for Nuclear-Grade Activated Carbon,'' into the 
Arkansas Nuclear One, Unit No. 1, Technical Specifications.
    Date of issuance: December 28, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 210.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.

[[Page 7689]]

    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12291). The application was renoticed on March 22, 2000 (65 FR 15378).
    The October 19, 2000, supplemental letter provided clarifying 
information and revised Bases pages that was within the scope of the 
application and did not change the associated no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 10, 1999, as supplemented 
by letters dated November 4, 1999, and October 12, 2000.
    Brief description of amendment: By letter dated June 10, 1999, 
FirstEnergy submitted its response for Davis-Besse Nuclear Power 
Station to the actions requested in Generic Letter (GL) 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated June 
3, 1999. By letter dated November 4, 1999, FirstEnergy requested 
changes to the Technical Specifications (TS) sections 3/4.6.4.4, 
``Hydrogen Purge System (HPS),'' 3/4.6.5.1, ``Shield Building Emergency 
Ventilation System (SBEVS),'' 3/4.7.6.1, ``Control Room Emergency 
Ventilation System (CREVS),'' and 6.0, ``Administrative Controls,'' for 
Davis-Besse Nuclear Power Station. FirstEnergy proposes adoption of a 
Ventilation Filter Testing Program (VFTP) in TS section 6.0--
Administrative Control and removal of the specific ventilation filter 
testing requirements from the plant's Surveillance Requirements of TS 
sections 3/4.6.4.4, 3/4.6.5.1, and 3/4.7.6.1. By letter dated October 
12, 2000, FirstEnergy provided additional information regarding 
relative humidity in the control room. The proposed changes would 
revise the TS surveillance testing of the safety related ventilation 
system charcoal to meet the requested actions of GL 99-02.
    Date of issuance: January 11, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 244.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73091).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated January 11, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 28, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.7.6.1, ``Plant Systems--Control Room Emergency 
Ventilation System,'' to establish actions to be taken for an 
inoperable control room ventilation system due to a degraded control 
room boundary (CRB). This revision approves changes that would allow up 
to 24 hours to restore the CRB to operable status when two control room 
ventilation system trains are inoperable due to an inoperable CRB in 
MODES 1, 2, 3, and 4. In addition, a Limiting Condition for Operation 
note would be added to allow the CRB to be opened intermittently under 
administrative controls without affecting control room ventilation 
system operability. Various other editorial changes have been made to 
reflect the revised TS. The applicable TS Bases have been revised to 
document the TS changes and to provide supporting information.
    Date of issuance: January 2, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 254.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46010).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 2, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: November 10, 1999, as 
supplemented October 3, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.7.c, to commit to the American Society for 
Testing and Materials D3803-1989 test protocol for the ventilation 
filter testing program. The changes are consistent with Nuclear 
Regulatory Commission (NRC) Generic Letter 99-02.
    Date of issuance: December 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 235.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1924).
    The supplemental information in the October 3, 2000, letter 
contained clarifying information and did not change the initial no 
significant hazards consideration determination and did not expand the 
scope of the original Federal Register notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
December 27, 2000.
    No significant hazards consideration comments received: No.

PECO Energy Company, PSEG Nuclear LLC, Delmarva Power and Light 
Company, and Atlantic City Electric Company, Docket Nos. 50-277 and 50-
278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, York County, 
Pennsylvania

    Date of application for amendments: October 10, 2000.
    Brief description of amendments: The amendments revised the 
licenses for Peach Bottom Units 2 and 3 to remove Delmarva Power and 
Light Company as a licensee, in conjunction with the transfer of the 
minority ownership interests of Delmarva Power and Light Company to the 
majority owners, PECO Energy Company and PSEG Nuclear LLC.
    Date of issuance: December 29, 2000.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendments Nos.: 238 & 241.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the License.
    Date of initial notice in Federal Register: November 27, 2000 (65 
FR 70740).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 27, 2000.

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 application for amendments: December 20, 1999, as 
supplemented

[[Page 7690]]

February 11, February 25, and October 10, 2000.
    Brief description of amendments: The amendments revised Facility 
Operating Licenses DPR-70 and DPR-75 to reflect changes related to the 
transfer of the license for the Salem Nuclear Generating Station, Unit 
Nos. 1 and 2, to the extent held by Delmarva Power and Light Company, 
to PSEG Nuclear Limited Liability Company.
    Date of issuance: December 29, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 240 and 221.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the License.
    Date of initial notice in Federal Register: February 18, 2000 (65 
FR 8452). The February 11, February 25, and October 10, 2000, 
supplements did not expand the scope of the original application with 
respect to both the proposed transfer action and the proposed amendment 
action as initially noticed in the Federal Register. No hearing 
requests or comments were received. In addition, the submittal did not 
affect the applicability of the Commission's generic no significant 
hazards consideration determination set forth in 10 CFR 2.1315.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 21, 2000.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: October 13, 1999, as 
supplemented by letter dated June 1, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications to permit relaxation of allowed bypass test 
times for Limiting Conditions for Operations (LCO) 3.3.1, ``Reactor 
Trip System Instrumentation'', and LCO 3.3.2, ``Engineered Safety 
Feature Actuation System Instrumentations''. These changes specifically 
revise the completion times from 6 hours to 72 hours for inoperable 
analog instruments, increase bypass times from 6 hours to 12 hours for 
surveillance testing of analog channels, and increase completion times 
from 6 hours to 24 hours for an inoperable logic cabinet or master and 
slave relays.
    Date of issuance: December 22, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 116 and 94.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46016).
    The supplemental letter dated June 1, 2000, provided clarifying 
information that did not change the scope of the October 13, 1999, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 22, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 17th day of January 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
 Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-1987 Filed 1-23-01; 8:45 am]
BILLING CODE 7590-01-P