[Federal Register Volume 66, Number 124 (Wednesday, June 27, 2001)]
[Notices]
[Pages 34278-34293]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-15818]


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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 June 4 through June 15, 2001. The last 
biweekly notice was published on June 12, 2001 (66 FR 31700).

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

[[Page 34279]]

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 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 July 27, 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 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: Rulemaking and 
Adjudications 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

[[Page 34280]]

Public Document Room, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Assess and Management 
Systems (ADAMS) Public Electronic Reading Room on the internet at the 
NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC Public Document room (PDR) 
Reference staff at 1-800-397-4209, 304-415-4737 or by email to 
<a href="/cdn-cgi/l/email-protection#04746076446a76672a636b72"><span class="__cf_email__" data-cfemail="f5859187b59b8796db929a83">[email protected]</span></a>.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: June 4, 2001.
    Description of amendments request: The proposed license amendments 
revise, from 2 hours to 6 hours, the time period in Surveillance 
Requirement 3.6.1.6.1 for verifying that each suppression chamber-to-
drywell vacuum breaker is closed after any discharge of steam to the 
suppression chamber from any source. In conjunction with this change, 
the Completion Time associated with Required Action B.1 for closing an 
open vacuum breaker is being revised from 8 hours to 4 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes provide additional time to verify that each 
vacuum breaker is closed and reduce the time allowed for closing an 
open vacuum breaker. The safety functions of the suppression 
chamber-to-drywell vacuum breaker valves are to relieve vacuum in 
the drywell following a postulated loss-of-coolant accident and to 
remain closed, except when the vacuum breakers are performing their 
intended design function, in order to ensure that no excessive 
bypass leakage occurs from the drywell to the suppression chamber. 
With a vacuum breaker not closed, communication between the drywell 
and suppression chamber airspaces could occur and, if a loss-of-
coolant accident were to occur, there would be the potential for 
primary containment overpressurization due [to] steam leakage from 
the drywell to the suppression chamber without quenching. The vacuum 
breakers do not perform a safety function that initiates, or alters 
initiation of, an accident previously evaluated. Rather, the vacuum 
breakers function to mitigate the consequences of certain design 
basis accidents. Therefore, the proposed changes do not involve an 
increase in the probability of an accident previously evaluated or 
the method of performing their safety functions.
    As noted above, the vacuum breakers function to mitigate the 
consequences of certain design basis accidents. The proposed changes 
to the Surveillance Requirement and Completion Time provide 
additional time to verify that each vacuum breaker is closed and 
reduce the time allowed for closing an open vacuum breaker; however, 
the proposed changes do not alter the safety functions of the vacuum 
breakers. When performing the surveillance to verify each vacuum 
breaker is closed, the expected result is the verification that the 
component is indeed closed. However, if this surveillance result is 
not obtained, the Technical Specifications limit the time allowed to 
close the vacuum breaker. Additional time is being provided to 
verify that each vacuum breaker is closed; however, the overall time 
allowed for closing and verifying closure of a vacuum breaker is not 
being increased. Since the overall time to take action for an open 
vacuum breaker has not been increased, the proposed changes do not 
involve an increase in the consequences of an accident previously 
evaluated.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The suppression chamber-to-drywell vacuum breakers are not an 
initiator of any design basis accident. Rather, the safety functions 
of the vacuum breaker valves are to relieve vacuum in the drywell 
following a loss-of-coolant accident and to remain closed when not 
relieving vacuum to ensure that no excessive bypass leakage occurs 
from the drywell to the suppression chamber. Neither safety function 
of these vacuum breakers is altered by the proposed changes. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed changes will not affect the ability of the 
suppression chamber-to-drywell vacuum breakers to perform their 
safety functions. Rather, as previously stated, the proposed changes 
provide additional time to verify that each vacuum breaker is closed 
and reduce the time allowed for closing an open or inoperable vacuum 
breaker. As a result, the overall time for taking action for an open 
vacuum breaker is unchanged. The vacuum breakers will continue to be 
verified closed every 14 days, as part of a required functional test 
of the vacuum breaker every 31 days, and following any activity 
involving the discharge of steam to the suppression chamber. If a 
vacuum breaker is found to be open and cannot be closed as required, 
plant shutdown will continue to be required within the same time 
requirements as currently specified in the Technical Specifications. 
Current Technical Specifications allow up to 10 hours to close an 
open vacuum breaker (i.e., 2 hours to perform the surveillance to 
verify vacuum breaker closure and, if necessary, 8 hours to close 
the vacuum breaker). The proposed change maintains the 10 hour limit 
by reducing the time to 4 hours to close an open or inoperable 
vacuum breaker while increasing the time to 6 hours to complete the 
surveillance to verify vacuum breaker closure. Thus, on this basis, 
the proposed license amendments will not change overall plant risk 
and 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Patrick M. Madden, Acting.

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: May 18, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.9.4 ``Containment Building 
Penetrations'' and the associated Bases to permit containment building 
penetrations to remain open, under administrative controls, during core 
alterations or the movement of irradiated fuel within the containment. 
Specifically, the licensee proposes: (1) Incorporating an alternate 
source term methodology in the fuel handling accident analysis; (2) 
revising TS 3.9.4 to remove portions of a note restricting the 
applicability of administrative controls with respect to containment 
penetrations; and (3) including the use of administrative controls on 
the equipment hatch and other penetrations that provide access from 
containment atmosphere to outside atmosphere.
    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 proposed changes modify TS requirements previously reviewed 
and

[[Page 34281]]

approved by the NRC in improved Technical Specifications (ITS) and 
changes to ITS as described in TSTF [Technical Specification Task 
Force]-312. An alternate source term calculation has been performed 
for the HNP [Harris Nuclear Plant] that demonstrates that dose 
consequences remain below limits specified in NRC Regulatory Guide 
1.183 and 10 CFR 50.67. The proposed change does not modify the 
design or operation of equipment used to move spent fuel or to 
perform core alterations[.]
    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.
    Containment penetrations are designed to form part of the 
containment pressure boundary. The proposed change provides for 
administrative controls and operating restrictions for containment 
penetrations consistent with guidance approved by the NRC staff. 
Containment penetrations are not an accident initiating system as 
described in the Final Safety Analysis Report [FSAR]. The proposed 
change does not affect other Structures, Systems, or Components. The 
operation and design of containment penetrations in operational 
modes 1-4 will not be affected by this proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes modify required Actions and Surveillance 
Requirements previously reviewed and approved by the NRC in improved 
Technical Specifications (ITS) and changes to ITS, TSTF-312. 
Additionally, the implementation of the alternate source term 
methodology is consistent with NRC Regulatory Guide 1.183. The 
proposed change to containment penetrations does not significantly 
affect any of the parameters that relate to the margin of safety as 
described in the Bases of the TS or the FSAR. Accordingly, NRC 
Acceptance Limits are not significantly affected by this change.
    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: Patrick M. Madden, Acting.
    Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina
    Date of amendment request: March 9, 2001.
    Description of amendment request: The amendment will revise the 
cold leg elbow tap flow coefficients used in the determination of 
Reactor Coolant System (RCS) flow rate at Catawba Nuclear Station, Unit 
2. No changes in Technical Specification are necessary for this 
amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
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.

First Standard

    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. No component modification, system realignment, or change 
in operating procedure will occur which could affect the probability 
of any accident or transient. The revised cold leg elbow tap flow 
coefficients will not change the probability of actuation of any 
Engineered Safeguards Feature or other device. The actual Unit 2 RCS 
flow rate will not change. Therefore, the consequences of previously 
analyzed accidents will not change as a result of the revised flow 
coefficients.

Second Standard

    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No component modification or system realignment will 
occur which could create the possibility of a new event not 
previously considered. No change to any methods of plant operation 
will be required. The elbow taps are already in place, and are 
presently being used to monitor flow for Reactor Protection System 
purposes. They will not initiate any new events.

Third Standard

    The proposed amendment will not involve a significant reduction 
in a margin of safety. The removal of some of the excess flow 
margin, which was introduced by the hot leg streaming flow penalties 
in later calorimetrics, will allow additional operating margin 
between the indicated flow and the Technical Specification minimum 
measured flow limit. The proposed changes in the cold leg elbow tap 
flow coefficients will continue to be conservative with respect to 
the analytical model flow predictions, since the proposed 
coefficients will continue to contain some hot leg streaming 
penalties from the calorimetric determined coefficients used in the 
average.
    An increase in the RCS flow indication of approximately 1.0% 
will increase the margin to a reactor trip on low flow but will not 
adversely affect the plant response to low flow transients. Current 
UFSAR Chapter 15 transients that would be expected to cause a 
reactor trip on the RCS low flow trip setpoint are Partial Loss of 
Reactor Coolant Flow, Reactor Coolant Pump Shaft Seizure and Reactor 
Coolant Pump Shaft break transients. Three reactor trip functions 
provide protection for these transients, RCS low flow reactor trip, 
RCP undervoltage reactor trip and RCP underfrequency reactor trip. 
The transient analyses of these events assume the reactor is tripped 
on the low flow reactor trip setpoint. This is conservative and 
produces a more severe transient response since a reactor trip on 
undervoltage or underfrequency would normally be expected to trip 
the reactor sooner and therefore reduce the severity of these 
transients.
    The RCS low flow reactor trip is currently set at 91% of the 
Technical Specification minimum measured flow of 390,000 gpm. The 
setpoint will not be revised as a result of this change, which means 
the transients relying on this function will behave in the same 
manner with the reactor trips occurring at essentially the same 
conditions as previously analyzed. Therefore, any small increase in 
the reactor trip margin gained by the small increase in the 
indicated RCS flow will not adversely affect the plant response 
during these low flow events.
    Based upon the preceding discussion, Duke Energy has concluded 
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: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: May 23, 2001.

[[Page 34282]]

    Description of amendment request: The amendment request proposes a 
change to the minimum critical power ratio safety limit (SLMCPR) and 
changes to the references for the analytical methods used to determine 
the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. 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 Minimum Critical Power Ratio (MCPR) safety limit is defined 
in the Bases to Technical Specification [TS] 2.1.1 as that limit 
which ``ensures that during normal operation and during AOOs 
[Anticipated Operational Occurrences], at least 99.9% of the fuel 
rods in the core do not experience transition boiling.'' The MCPR 
safety limit satisfies the requirements of General Design Criterion 
10 of Appendix A to 10 CFR [Part] 50 regarding acceptable fuel 
design limits. The MCPR safety limit is re-evaluated for each reload 
using NRC [Nuclear Regulatory Commission]-approved methodologies. 
The analyses for RBS [River Bend Station] Cycle 11 have concluded 
that a two-loop MCPR safety limit of 1.08, based on the application 
of Framatome ANP Richland, Inc.'s [FRA-ANP] [(proprietary)] NRC-
approved MCPR safety limit methodology, will ensure that this 
acceptance criterion is met. For single-loop operation, a MCPR 
safety limit of 1.10, also ensures that this acceptance criterion is 
met.
    In addition to the MCPR safety limit, core operating limits are 
established to support the Technical Specification 3.2 requirements 
which ensure that the fuel design limits are not exceeded during any 
conditions of normal operation or in the event of any anticipated 
operational occurences (AOO). The methods used to determine the core 
operating limits for each operating cycle are based on methods 
previously found acceptable by the NRC and listed in TS section 
5.6.5. A change to TS section 5.6.5 is requested to include the FRA-
ANP methods in the list of NRC approved methods applicable to RBS. 
These NRC approved methods will continue to ensure that acceptable 
operating limits are established to protect the fuel cladding 
integrity during normal operation and in the event of an AOO.
    The requested Technical Specification changes do not involve any 
plant modifications or operational changes that could affect system 
reliability or performance or that could affect the probability of 
operator error. The requested changes do not affect any postulated 
accident precursors, do not affect any accident mitigating systems, 
and do not introduce any new accident initiation mechanisms.
    Therefore, these changes to the Minimum Critical Power Ration 
(MCPR) safety limit and to the list of methods used to determine the 
core operating limits do 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 ATRIUM-10 fuel to be used in Cycle 11 is of a design 
compatible with the co-resident GE-11. Therefore, the introduction 
of ATRIUM-10 fuel into the Cycle 11 core will not create the 
possibility of a new or different kind of accident. The proposed 
changes do not involve any new modes of operation, any changes to 
setpoints, or any plant modifications. The proposed revised MCPR 
safety limits have accounted for the mixed fuel core and have been 
shown to be acceptable for Cycle 11 operation. Compliance with the 
criterion for incipient boiling transition continues to be ensured. 
The core operating limits will continue to be developed using NRC 
approved methods which also account for the mixed fuel core design. 
The proposed MCPR safety limits or methods for establishing the core 
operating limits do not result in the creation of any new precursors 
to an accident.
    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 MCPR safety limits have been evaluated in accordance with 
Framatome ANP Richland, Inc.'s NRC-approved cycle-specific safety 
limit methodology to ensure that during normal operation and during 
Anticipated Operational Occurrences (AOO's) at least 99.9% of the 
fuel rods in the core are not expected to experience transition 
boiling. On this basis, the implementation of this Framatome ANP 
Richland, Inc. methodology does not involve a significant reduction 
in a margin of safety.
    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: May 11, 2001.
    Description of amendment request: This amendment revised the 
Technical Specifications to allow, on a one-time basis only, Entergy 
Nuclear Operations, Inc. to extend the allowed out-of-service time for 
the Residual Heat Removal Service Water (RHRSW) System from 7 days to 
11 days. This amendment is only applicable during installation of the 
modification 00-12 to the ``B'' RHRSW Strainer.
    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:

    Involve an increase in the probability or consequences of an 
accident previously evaluated.
    The CCDP [Conditional Core Damage Probability] due to this 
proposed change is calculated to be 4.33 E-8 (assuming no-risk 
significant SSC maintenance), which falls below the threshold 
probability of 1 E-6 for risk significance of temporary changes to 
the plant configuration in the EPRI PSA Applications Guide 
(Reference 2). The ICLERP [incremental conditional large early 
release probability] is calculated to be 8.85 E-8, which falls below 
the threshold probability of 1 E-7 for risk significance per 
Reference 2 [see application dated May 11, 2001].
    This proposed change does not increase the consequences of an 
accident previously evaluated because all relevant accidents (LOCA) 
[loss-of-coolant accident] would result in the transfer of decay 
heat to the suppression pool. For this scenario, the same compliment 
of equipment will be available to achieve and maintain cold shutdown 
as is required by the current TS LCO [limiting condition for 
operation].
    Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not physically alter the plant. As 
such, no new or different types of equipment will be installed. The 
new design for the RHRSW strainer packing gland will be evaluated 
under a separate 10 CFR 50.59 evaluation and is considered to be 
functionally equivalent for the purposes of this one-time-only 
proposed TS change.
    The connection and use of a temporary hose for achieving limited 
containment heat removal in the event the ``A'' division of RHRSW is 
rendered inoperable for some reason is a contingency plan that is 
already addressed by current plant procedures.
    Involve a significant reduction in a margin of safety.
    The CCDP due to this proposed change is calculated to be 4.33 E-
8 (assuming no-risk significant SSC maintenance). This value falls 
below the threshold probability of 1 E-6 for risk significance of 
temporary changes to the plant configuration in the EPRI PSA 
Applications Guide (Reference 2). The CLERP is calculated to be 8.85 
E-8, which falls below the threshold probability of 1 E-7 for risk 
significance per Reference 2.
    The consequences of a postulated accident occurring during the 
extended allowable out-

[[Page 34283]]

of-service time are bounded by existing analyses, therefore, there 
is no significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: May 22, 2001.
    Description of amendment request: The proposed change to Technical 
Specification (TS) 3/4.7.1.2, Emergency Feedwater (EFW) System expands 
and clarifies the current TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The administrative and more restrictive changes will not affect 
the assumptions, design parameters, or results of any accident 
previously evaluated. The accident mitigation features of the plant 
are not affected by these proposed changes. The proposed changes do 
not add or modify any existing equipment. The administrative change 
to test EFW pumps pursuant to the Inservice Test Program will ensure 
the EFW pumps are tested against the more restrictive of the data 
points required by either the safety analysis or the Inservice Test 
Program. Therefore, the proposed administrative changes do not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    The less restrictive changes (allowing 7 days for an inoperable 
pump due to an inoperable steam supply, allowing 24 hours for an 
inoperable steam supply and one inoperable motor driven EFW pump, 
allowing 72 hours for two inoperable motor driven EFW pumps, 
performing Surveillance Requirements during other than shutdown 
conditions, allowing the use of actual actuation signals in addition 
to test signals, and delaying the requirement to complete 
Surveillance Requirement ``d'' to just prior to Mode 2) will not 
affect the assumptions, design parameters, or results of any 
accident previously evaluated. The accident mitigation features of 
the plant are not affected by these proposed changes. The proposed 
changes do not add or modify any existing equipment. Therefore, the 
proposed less restrictive changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    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 proposed changes do not alter the design or configuration of 
the plant. There has been no physical change to plant systems, 
structures, or components. The proposed changes will not reduce the 
ability of any of the safety-related equipment required to mitigate 
Anticipated Operational Occurrences or accidents.
    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 proposed change to the LCO [Limiting Conditions for 
Operation] requiring three pumps and two flow paths be OPERABLE 
maintains the functionality of the EFW such that it is capable of 
performing its design function as assumed in the Final Safety 
Analysis Report. If the functionality of the system is not 
maintained, Technical Specifications require ACTIONs be taken, 
within specified time limitations, to restore EFW to OPERABLE status 
or shutdown the reactor. This action is consistent with the existing 
Technical Specifications and NUREG-1432.
    The allowed outage time for one inoperable steam supply has been 
increased from 72 hours to 7 days in accordance with NUREG-1432. 
This is acceptable due to the redundant OPERABLE steam supply, the 
availability of redundant OPERABLE motor-driven EFW pumps, and the 
low probability of an event requiring the inoperable steam supply. 
This change is consistent with NUREG-1432 and has therefore been 
previously approved by the NRC [Nuclear Regulatory Commission].
    The ACTION for an inoperable steam supply to the turbine-driven 
EFW pump steam turbine concurrent with one motor-driven EFW pump 
being inoperable will allow a 24 hour completion time. This change 
is acceptable based on the ability of the system to cool the reactor 
coolant system to shutdown cooling entry conditions following a loss 
of normal feedwater. The 24 hour completion time is reasonable based 
on the redundant OPERABLE steam supply to the turbine-driven EFW 
pump steam turbine, the OPERABLE motor-driven EFW pump, and the low 
probability of an event requiring the inoperable steam supply to the 
turbine-driven EFW pump.
    The ACTION for an inoperable steam supply to the turbine-driven 
EFW pump steam turbine concurrent with both motor-driven EFW pumps 
being inoperable as proposed requires a unit shutdown be initiated 
immediately. This change is appropriate due to the seriousness of 
the condition and is acceptable due to the ability of the EFW system 
to support the unit shut down.
    The ACTION for the EFW system inoperable for reasons other than 
those described in ACTION (a), (b), or (c) and able to deliver at 
least 100% flow to either steam generator as proposed will allow a 
72 hour completion time. This change is acceptable based on the 
ability of the system to cool the RCS [Reactor Coolant System] to 
SDC [Shutdown Cooling] entry conditions following a design basis 
accident assuming no single active failure.
    The ACTION for the EFW system inoperable for reasons other than 
those described in ACTION (a), (b), or (c) and able to deliver at 
least 100% combined flow to the steam generators as proposed 
requires a unit shutdown be initiated immediately. This change is 
appropriate due to the seriousness of the condition and is 
acceptable due to the ability of the EFW system to support the unit 
shut down.
    The ACTION for the EFW system inoperable and unable to deliver 
at least 100% flow to the steam generators as proposed requires 
immediate action be taken to restore the ability to deliver at least 
100% flow to the steam generators. The unit is in a seriously 
degraded condition in that the EFW system is unable to support a 
unit shutdown. This change is consistent with the intent of the 
current EFW Technical Specification and NUREG-1432.
    Testing pursuant to Specification 4.0.5 (Inservice Testing 
Program) as proposed for Surveillance Requirement `b' will ensure 
the EFW pumps are tested against the more restrictive of the data 
points required by either the safety analysis or ASME [American 
Society of Mechanical Engineers] Section XI.
    The remaining changes to the EFW Technical Specification are 
consistent (other than format) with NUREG-1432 and have therefore 
been previously approved by the NRC.
    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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 28, 2001
    Description of amendment request: The proposed amendments would

[[Page 34284]]

revise the Technical Specifications to eliminate the requirement for at 
least one person qualified to stand watch to be present in the control 
room when nuclear fuel is stored in the spent fuel pool.
    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 Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The Defueled Safety Analysis Report (DSAR) identifies three 
categories of events: spent fuel pool events (i.e., operational 
occurrences), fuel handling accidents in the fuel building, and 
radioactive waste handling accidents. There are no active controls 
in the control room that affect spent fuel pool equipment, or the 
handling of fuel or radioactive waste. Actions to mitigate the 
consequences of these events are taken outside the control room. 
Emergency response is not adversely affected by this proposed change 
because the control room is still available to the emergency 
response team and communication capability and timeliness will not 
be affected. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The configuration, operation and accident response of the 
systems, structures or components that support safe storage of the 
spent fuel are unchanged by the proposed TS change. Current site 
surveillance requirements ensure frequent and adequate monitoring of 
system and component functionality. Systems in the Spent Fuel 
Nuclear Island will continue to be operated in accordance with 
current design requirements and no new components or system 
interactions have been identified. No new accident scenarios, 
failure mechanisms or limiting single failures are introduced as a 
result of the proposed change. The proposed TS change does not have 
an adverse affect on any system related to safe storage of spent 
fuel. Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    All design basis accident acceptance criteria will continue to 
be met. The margin of safety relative to the cooling of the spent 
fuel is unaffected by the proposed change as the SFP [spent fuel 
pool] parameters will continue to be monitored at the same frequency 
that they are monitored now. The ability of the shift crew to 
respond to abnormal or accident conditions is unaffected by the 
proposed change since all controls are located in the fuel building 
and any necessary communication will be handled by the DERO 
[Defueled Emergency Response Organization]. Therefore, it is 
concluded that the proposed TS 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Robert Helfrich, Senior Counsel, 
Nuclear, Mid-West Regional Operating Group, Exelon Generation Company, 
LLC, 1400 Opus Place, Suite 900, Downers Grove, Illinois 60515.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 15, 2001.
    Description of amendment request: The proposed amendment would 
revise refueling operation Technical Specification (TS) requirements 
for containment equipment hatch cover closure during core alterations 
and during movement of irradiated fuel both inside containment and in 
the spent fuel pool or cask pit. The proposed change would allow the 
containment equipment hatch cover to be off during core alterations and 
movement of irradiated fuel provided the Emergency Ventilation System 
is operable with the ability to filter any radioactive release. The 
proposed changes involve TS 3/4.9.4, Refueling Operations -Containment 
Penetrations, and TS 3/4.9.12, Refueling Operations--Storage Pool 
Ventilation, and associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed changes. The amendment application proposes to 
revise DBNPS TS 3/4.9.4, Refueling Operations--Containment 
Penetrations, and its associated Bases, and TS 3/4.9.12, Refueling 
Operations--Storage Pool Ventilation, and its associated Bases. The 
proposed changes would provide for access to the containment through 
the containment equipment hatch during core alterations and movement 
of irradiated fuel, provided that an Emergency Ventilation System is 
operable with the ability to filter any radioactivity release 
through the containment equipment hatch. The proposed changes would 
also permit relying on the closing the containment personnel air 
lock by a designated individual to establish the negative pressure 
boundary for the Emergency Ventilation System servicing the storage 
pool. The use of a designated individual to close the containment 
personnel airlock is currently permitted by TS 3.9.4 for meeting 
containment closure requirements. Neither the containment equipment 
hatch nor the Emergency Ventilation System contributes to the 
initiation of any accident described in the DBNPS Updated Safety 
Analysis Report.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no equipment, accident 
conditions, or assumptions are affected which could lead to a 
significant increase in radiological consequences. The approved 
analysis for the fuel handling accident inside containment does not 
take credit for containment closure or Emergency Ventilation System 
filtering. This analysis results in a maximum calculated offsite 
does well within the limits of 10 CFR 100.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new or 
different accident initiators are introduced by these proposed means 
to mitigate the consequences of an accident.
    3. Not involve a significant reduction in a margin of safety 
because there are no changes to the initial conditions contributing 
to accident severity or the resulting consequences. Consequently, 
there are no significant reductions in a margin of safety.
    On the basis of the above, the Davis-Besse Nuclear Power Station 
has determined that the License Amendment Request does not involve a 
significant hazards consideration. As this License Amendment Request 
concerns a proposed change to the Technical Specifications that must 
be reviewed by the Nuclear Regulatory Commission, this License 
Amendment Request does not constitute an unreviewed safety question.

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

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

    Date of amendment request: May 14, 2001.

[[Page 34285]]

    Description of amendment request: The proposed amendments would 
delete Technical Specifications (TS) Figures 5.1-1, ``Site Area Map,'' 
and 5.1-2, ``Plant Area Map,'' and would replace TS 5.1, ``Site,'' with 
a site location description. Conforming changes are requested to delete 
TS 5.1.1, ``Exclusion Area,'' TS 5.1.2, ``Low Population Zone,'' and TS 
5.1.3, ``Map Defining Unrestricted Areas and Site Boundary for 
Radioactive Gaseous and Liquid Effluents,'' from TS 5.1 and the TS 
Index. These changes conform to NUREG-1431, Rev. 1, Improved Standard 
TS for Westinghouse Plants, and the requirements of 10 CFR 50.36 
(c)(4).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.

    The proposed amendments are administrative in nature, removing 
sections and maps from the TS, which are located in other documents 
previously approved by NRC. These amendments will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated because they do not affect assumptions 
contained in plant safety analyses, the physical design and/or 
operation of the plant, nor do they affect TS that preserve safety 
analysis assumptions. Therefore, the proposed changes do not affect 
the probability or consequences of accidents previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes to the TS are administrative in nature and 
can not create the possibility of a new or different kind of 
accident from any previously evaluated since the proposed amendments 
will not change the physical plant or the modes of plant operation 
defined in the facility operating license. No new failure mode is 
introduced due to the administrative changes since the proposed 
changes do not involve the addition or modification of equipment, 
nor do they alter the design or operation of affected plant systems, 
structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes are administrative in nature and do not 
affect operating limits or functional capabilities of plant systems, 
structures and components. The addition of a site location 
description to the TS adds geographical information to the TS. 
Elimination of site and plant area maps from the TS would have no 
effect on margin of safety as they are located in other controlled 
plant documents. Thus, the changes proposed would not involve a 
significant reduction in margin of safety of the facility.

    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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Patrick M. Madden (Acting).

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

    Date of amendment request: February 15, 2001.
    Description of amendment request: The proposed amendment to the 
Cooper Nuclear Station (CNS) Operating License (OL) DPR-46 would (1) 
delete OL Condition 2.D, Additional Conditions for Protection of the 
Environment, and (2) remove the depiction of railroad tracks in 
Technical Specifications (TS) Figure 4.1-1, Site and Exclusion Area 
Boundaries and Low Population Zone.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    OL [Operating License] Condition 2.D has become obsolete based 
upon it being satisfied or superceded by amendments to the FSAR 
[Final Safety Analysis Report] and OL. The previous FSAR and OL 
amendments which made it obsolete were reviewed and approved based 
on their individual Unreviewed Safety Question (USQ) evaluations or 
no significant hazards considerations. Since this proposed change 
does not physically alter any plant equipment or operating 
limitations, it therefore does not impact any previously evaluated 
accident initiator, nor change mitigating systems or features or 
operating limitations for accidents previously evaluated in the 
Updated Safety Analysis Report (USAR). Thus, it does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. This is an administrative change.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This proposed change is administrative in nature. It does not 
involve a physical alteration of the plant. No new or different 
equipment is being installed, and no installed equipment is being 
operated in a new or different manner. No setpoints for parameters 
which initiate protective or mitigative action are being changed. As 
a result, no new failure modes are being introduced. There are no 
changes in the procedures or methods governing normal plant 
operation, nor are the procedures utilized to respond to plant 
transients altered as a result of this administrative change. This 
change does not impose any new or different requirements or 
eliminate any existing requirements. In addition, the change does 
not alter assumptions made in the safety analysis, nor does it 
impact the licensing basis. Therefore, the changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    This proposed change is administrative in nature. It does not 
alter any accident analysis assumptions, conditions, or methodology. 
Since this proposed change does not physically alter plant systems, 
structures or components (SSC's), change mitigating systems, 
features, operating limitations, nor revise accident analysis 
assumptions, conditions or methodology, it does not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: October 19, 2000, as supplemented March 
23 and April 9, 2001.
    Description of amendment request: The proposed amendment would 
authorize the licensee to change the licensing basis to utilize the 
full scope of an alternative radiological source term for accidents as 
described in NUREG-1465, ``Accident Source Terms for Light-Water 
Nuclear Power Plants,'' and change the Technical Specifications

[[Page 34286]]

to implement various assumptions in the Alternative Source Term 
analyses. The portion of this amendment request regarding operability 
requirements during core alterations and while moving irradiated fuel 
assemblies within the secondary containment, and which provided for 
selective application of the Alternative Source Term to the design-
basis fuel handling accident was previously evaluated and issued as 
Amendment No. 237 on April 16, 2001.
    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 Alternative Source Term and those plant systems affected by 
implementing the setpoints and action levels specified in the 
analyses are not assumed to initiate design basis accidents. The 
Alternative Source Term does not affect the design or operation of 
the facility; rather, once the occurrence of an accident has been 
postulated the new source term is an input to evaluate the 
consequence. The implementation of the Alternative Source Term has 
been evaluated in revisions to the analyses of the limiting design 
bases accidents at DAEC [Duane Arnold Energy Center]. Based on the 
results of these analyses, it has been demonstrated that, with the 
requested changes, the dose consequences of these limiting events 
are within the regulatory guidance provided by the NRC for use with 
the Alternative Source Term. This guidance is presented in NUREG 
1465, 10 CFR 50.67, associated Regulatory Guide 1.183, and Standard 
Review Plan (SRP) Section, 15.0.1. Since secondary containment 
operability is not assumed for the fuel handling accident (FHA), the 
consequences of eliminating the requirements for secondary 
containment operability, secondary containment isolation valves/
dampers, secondary containment instrumentation and the Standby Gas 
Treatment system during fuel movement or core alterations will not 
increase the effects of a FHA beyond those evaluated in the 
Alternative Source Term analysis. Therefore, the proposed changes do 
not significantly increase the probability or consequences of any 
previously evaluated accident.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The Alternative Source Term and those plant systems affected by 
implementing the setpoints and action levels specified in the 
analyses do not initiate design basis 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 amendment will not involve a significant 
reduction in a margin of safety.
    The changes proposed are associated with the implementation of a 
new licensing basis for DAEC. Approval of the basis change from the 
original source term developed in accordance with TID-14844 to a new 
alternative source term as described in NUREG-1465 is requested by 
this submittal. The results of the accident analyses revised in 
support of this submittal, and the requested Technical Specification 
changes, are subject to revised acceptance criteria. These analyses 
have been performed using conservative methodologies. Safety margins 
and analytical conservatisms have been evaluated and are satisfied. 
The analyzed events have been carefully selected and margin has been 
retained to ensure that the analyses adequately bound all postulated 
event scenarios. The dose consequences of these limiting events are 
within the acceptance criteria also found in the latest regulatory 
guidance. This guidance is presented in NUREG 1465, in the approved 
rulemaking for 10 CFR 50.67, and in the associated Regulatory Guide 
1.183.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries, as well as the 
control room, are within the corresponding regulatory limit. 
Specifically, the margin of safety for these accidents is considered 
to be that provided by meeting the applicable regulatory limit, 
which, for most events, is conservatively set below the 10 CFR 50.67 
limit. With respect to the control room personnel doses, the margin 
of safety (the difference between the 10 CFR 50.67 limits and the 
regulatory limit defined by 10 CFR50, Appendix A, Criterion 19 (GDC 
19)) continues to be satisfied.
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, they are 
considered to not result in a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    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: May 30, 2001.
    Description of amendment request: The proposed amendment would 
eliminate local suppression pool temperature limits from the Updated 
Safety Analysis Report as the basis for limiting suppression pool 
mechanical loads due to unstable steam condensation during safety 
relief valve actuations.
    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.
    Eliminating the Local Suppression Pool Temperature Limits 
(LSPTLs) will not introduce new equipment or new equipment methods 
of operation, and will not alter existing system relationships. 
LSPTLs are not an accident initiator and does [sic] not affect other 
accident initiators. The integrity of fission product barriers do 
not rely on LSPTLs since mechanical loads on containment will not be 
exceeded and ECCS [emergency core cooling system] operation in the 
event of an accident will not be adversely affected as demonstrated 
and approved in Reference 6 [letter from G. Holahan (NRC) to R. 
Pinelli (Boiling Water Reactor Owners Group), ``Transmittal of the 
Safety Evaluation of General Electric Co. Topical Reports; NEDO-
30832, Entitled 'Elimination of Limit on BWR Suppression Pool 
Temperature for SRV Discharge With Quenchers,' and NEDO-31695, 
Entitled 'BWR Suppression Pool Temperature Technical Specification 
Limits','' dated August 29, 1994].
    Therefore, the proposed amendment will not significantly 
increase the probability or the consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    Eliminating the LSPTLs will not introduce new equipment or new 
equipment methods of operation, and will not alter existing system 
relationships. Since containment integrity and ECCS operation will 
not be challenged, new or different kinds of accidents are not 
created.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Since LSPTLs are not required to limit mechanical loads on 
containment, the margin of safety associated with containment 
integrity is not significantly reduced. Since LSPTLs are not 
required to prevent steam binding of the ECCS pumps, the margin of 
safety associated with ECCS operation is not significantly reduced.
    Therefore, the proposed amendment will not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 34287]]

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.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1,

    Washington County, Nebraska
    Date of amendment request: May 15, 2001.
    Description of amendment request: The proposed changes would: (1) 
Replace the titles of Manager--Fort Calhoun Station and the Vice 
President with generic titles, (2) relocate the requirements for the 
Plant Review Committee (PRC) and the Safety Audit and Review Committee 
(SARC) to the Fort Calhoun Station (FCS) Quality Assurance Program, (3) 
relocate the requirements for procedure controls and records retention 
to the FCS Quality Assurance Program, (4) enhance and clarify the 
qualification and training requirements for individuals who perform 
licensed operator functions, (5) incorporate the Westinghouse/CENP 
definition of Azimuthal Power Tilt, and (6) eliminate specific mailing 
address and reporting requirements that are redundant to Title 10 of 
the Code of Federal Regulations (10 CFR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes: revise the FCS definition of Azimuthal 
Power Tilt, remove specific titles from the Technical 
Specifications, provide minor clarifications of the training 
requirements for plant staff, and indicate the change in title of 
the Licensed Senior Operator. This change also relocates the 
requirements for the Plant Review Committee (PRC) and the Safety 
Audit and Review Committee (SARC), procedure control, and records 
retention to the Fort Calhoun Station Quality Assurance Program as 
described in NRC Administrative Letter 95-06.
    The proposed change includes an update to the definition of 
Azimuthal Power Tilt and adds the bases for the definition of 
Azimuthal Power Tilt to the bases section of Section 2.10.4 as 
recommended in ABB Combustion Engineering (CE) Infobulletin Number 
97-07, dated December 31, 1997. As noted in the infobulletin, CE 
discovered a discrepancy in the definition for CE analog plants that 
use Combustion Engineering Core Operating Report (CECOR) for 
monitoring and surveillance purposes. Plants that use CECOR should 
use the same definition as the CE digital plants. This change will 
make the FCS definition and bases agree with the improved Standard 
Technical Specifications for CE digital plants, which have 
previously been approved by the NRC.
    The proposed change would allow the use of generic personnel 
titles as provided in ANSI/ANS 3.1 and NUREG-1432, ``Standard 
Technical Specifications Combustion Engineering Plants,'' in lieu of 
plant-specific personnel titles. This change does not eliminate any 
of the qualifications, responsibilities or requirements for these 
positions, since the plant-specific personnel titles are currently 
identified in licensee controlled documents such as the Updated 
Safety Analysis Report (USAR) or the Quality Assurance Program. For 
example, Section 12 of the Updated Safety Analysis Report describes 
the management structure and reporting responsibilities of OPPD and 
provides an organizational chart to determine the corporate officer 
with responsibility for overall plant nuclear safety from other 
corporate officers within OPPD. Therefore, changing the terminology 
within the Technical Specifications, indicating this reporting 
responsibility does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Changing the periodicity of review for staff overtime is also 
considered an administrative change. This includes a change of the 
title of the Supervisor--Operations to Manager--Shift Operations, 
Licensed Senior Operator to Control Room Supervisor, and crewman to 
crewmember. The change to the number of Senior Operator License 
present during Core Alterations and the associated note is also 
considered clarifying in nature and not a change of intent.
    The proposed change would update the qualification requirements 
for the Manager--Radiation Protection, the Shift Technical Advisors, 
and those individuals that perform the functions described in 10 CFR 
50.54(m) to Regulatory Guide 1.8, Revision 3, and ANSI/ANS 3.1-1993. 
In the March 1987 revision to 10 CFR Part 55, the NRC included the 
requirement that those facility licensees that have made a 
commitment that is less than that required by the new rules must 
conform to the new rules automatically. OPPD had previously 
considered that commitments made to comply with the requirements of 
NUREG-0737 and the standards applied through the Institute of 
Nuclear Power Operations (INPO) accreditation process were 
equivalent to the guidance provided in Regulatory Guide 1.8, 
Revision 3. The proposed change provides enhancement to the current 
requirements and clarifies the qualifications and training 
requirements for licensed personnel. This provides additional 
assurance that these personnel are properly trained and qualified 
for their positions and conforms with the guidance of NRC Regulatory 
Issues Summary 2001-01. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed change would relocate specific requirements for 
SARC, PRC, procedure control, and records retention to the Fort 
Calhoun Station Quality Assurance Program (Appendix A, of the FCS 
USAR). This proposed revision does not change or eliminate 
responsibilities or requirements for these programs. The management 
level and expertise of personnel who are PRC or SARC members is not 
being changed. The review of plant operations, procedures control, 
and record retention is still required to be in compliance with the 
Fort Calhoun Station Quality Assurance (QA) Program. Any changes in 
the QA Program which reduce the effectiveness of the program must be 
approved by the NRC in accordance with 10 CFR 50.54(a)(4). These 
changes meet the criteria as described in NRC Administrative Letter 
95-06. Therefore, the proposed relocation of these programs to the 
QA Program does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change would also remove the requirements 
prescribing specific submittal addresses, titles, and reporting 
periods. For example, the requirement to submit License[e] Event 
Reports within 30 days is replaced with a citation referencing 10 
CFR 50.73. This is in agreement with 10 CFR 50.73 and 10 CFR 
50.4(f). Additionally, an administrative requirement prescribing the 
submittal of a Special Maintenance Report is being deleted, as it is 
redundant to the requirements of 10 CFR 50.73. 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 organizational and administrative 
requirements contained within the Administrative Controls section of 
the TS. The proposed change to the definition of Azimuthal Power 
Tilt is as recommended in CE Infobulletin 97-07 for CE analog plants 
that use CECOR for monitoring and surveillance purposes and will 
have no affect on accidents previously evaluated. The proposed 
changes do not revise any equipment setpoints, change the manner in 
which any plant equipment is operated, or propose any new operating 
modes. Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes revise organizational and administrative 
requirements contained within the Administrative Controls section of 
the TS. The proposed change to the definition of Azimuthal Power 
Tilt has no affect on the margin of safety. The proposed changes do 
not revise any equipment setpoints, change the manner in which any 
plant equipment is

[[Page 34288]]

operated, or propose any new operating modes. 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: May 17, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to permit an increase in the 
allowable leak rate for the main steam isolation valves (MSIVs) and to 
delete the MSIV Sealing System (MSIVSS). These changes are based on the 
use of an alternate source term and the guidance provided in Regulatory 
Guide 1.183, ``Alternate Radiological Source Terms for Evaluating 
Design Basis Accidents at Nuclear Power Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's review is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    As described in Section 6.7 of the Hope Creek Updated Final 
Safety Analysis Report (UFSAR), the MSIVSS limits the leakage of 
fission products through the MSIVs following a design-basis accident 
large break Loss of Coolant Accident (LOCA). The system is manually 
actuated following a LOCA. The licensee has proposed to remove the 
MSIVSS from the plant and to delete the associated requirements from 
the TSs. In addition, the TSs would be revised to increase the 
allowable MSIV leak rate. The MSIVSS lines and main steamline drain 
valves that are connected to the main steam piping will be capped 
and welded closed to ensure primary containment integrity is 
maintained. The welding and post-weld examination procedures will be 
in accordance with the American Society of Mechanical Engineers 
Code, Section III requirements. The welded caps will be periodically 
tested as part of the Containment Integrated Leak Rate Test. MSIV 
leakage and operation of the MSIVSS do not affect the precursors for 
accidents analyzed in Chapter 15 of the Hope Creek UFSAR. In 
addition, the proposed changes do not adversely affect other 
structures, systems, or components important to safety. Therefore, 
there is no increase in the probability of occurrence of an accident 
previously evaluated as a result of the proposed changes.
    The licensee's submittal states that the radiological 
consequences associated with the proposed changes have been analyzed 
based on the results of revised offsite and control room operator 
dose calculations for a LOCA, which is the most limiting Hope Creek 
design-basis accident. The current design-basis analysis for the 
radiological consequences associated with a LOCA is shown in Hope 
Creek UFSAR Sections 6.4.7 and 15.6.5.5. The revised analysis was 
performed using an alternate source term in accordance with the 
requirements in 10 CFR 50.67 and the guidance in Regulatory Guide 
1.183. The dose calculations assess the effects of the proposed 
increase in allowable MSIV leak rate and take no credit for the 
MSIVSS. In addition, the calculations assume an unfiltered control 
room inleakage design-basis value that is higher than the current 
design basis value to address control room habitability issues 
associated with NEI 99-03. The revised analysis was performed in 
accordance with the current accepted methodology discussed in 
Regulatory Guide 1.183 and the radiological consequences were 
evaluated in terms of Total Effective Dose Equivalent (TEDE) dose as 
per the acceptance criteria specified in 10 CFR 50.67. The 
Regulatory Guide 1.183 methodology is not exactly comparable to the 
current Hope Creek design basis analysis which is in terms of whole 
body and thyroid doses. The results of the licensee's analysis 
associated with the proposed changes indicate that the post-LOCA 
doses will result in an increase in the dose exposures for the 
control room, the Exclusion Area Boundary (EAB), and the Low 
Population Zone (LPZ), compared to the current design basis 
analysis. However, the revised post-LOCA doses will remain below the 
TEDE dose acceptance criteria for the control room, EAB, and LPZ, as 
specified in 10 CFR 50.67. The methodology and guidance provided in 
Regulatory Guide 1.183 has been developed for the purpose of 
performing design basis radiological consequence analyses using an 
alternate source term such that meeting the 10 CFR 50.67 acceptance 
criteria demonstrates adequate protection of public health and 
safety. Therefore, the proposed changes do not involve a significant 
increase in the consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to increase the allowed MSIV leakage rate 
does not affect the operability the MSIVs and will not inhibit the 
capability of the MSIVs to perform their function of isolating the 
primary containment as assumed in the Hope Creek accident analyses 
in UFSAR Chapter 15. The proposed change to delete the MSIVSS does 
not introduce any new modes of plant operation and, as previously 
discussed, the design-basis LOCA analysis was reanalyzed without 
taking credit for the operation of MSIVSS. The affected main steam 
piping will be welded and/or capped closed to assure that the 
primary containment integrity, isolation, and leak testing 
capability are not compromised. Based on the above considerations, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    As previously discussed, the results of the licensee's analysis 
associated with the proposed changes indicate that the post-LOCA 
doses will result in an increase in the dose exposures for the 
control room, the EAB, and the LPZ, compared to the current design 
basis analysis. Since there will be an increase in dose exposure, 
the margin of safety will be decreased. However, the revised post-
LOCA doses will remain below the TEDE dose acceptance criteria for 
the control room, EAB, and LPZ, as specified in 10 CFR 50.67. 
Meeting the 10 CFR 50.67 acceptance criteria demonstrates adequate 
protection of public health and safety. An acceptable margin of 
safety is inherent in these acceptance criteria. Therefore, there is 
no significant reduction in the margin of safety as a result of the 
proposed changes.

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

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: March 5, 2001.
    Description of amendment request: The proposed amendment would (1) 
change the Security Plan provision that a member of the security force 
escort all vehicles, other than designated licensee vehicles, and to 
delete the related Security Training and Qualification Plan task, (2) 
change the requirement of the Security Plan that all areas of the 
protected area be illuminated to a minimum of 0.2 footcandle, and (3) 
change the frequency of protected area patrols in the Security Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or

[[Page 34289]]

consequences of an accident previously evaluated?
    The proposed changes involving security activities do not reduce 
the ability for the security organization to prevent radiological 
sabotage and therefore do not increase the probability or 
consequences of a radiological release previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve functions of the security 
organization concerning vehicle control, protected area 
illumination, and protected area patrol frequency. Analysis of the 
proposed changes has not indicated nor identified a new or different 
kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Analysis of the proposed changes show that they affect only the 
functions of the Security organization and have no impact upon nor 
cause a significant reduction in margin of safety for plant 
operation. The failure points of key safety parameters are not 
affected.

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

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

    Date of amendment request: May 30, 2001 (ULNRC-04481).
    Description of amendment request: The proposed amendment changes 
the technical specifications to remove the phrase ``and the charging 
flow control valve full open'' from Limiting Condition for Operation 
3.5.5, Required Action A.1, and Surveillance Requirement 3.5.5.1 for 
the reactor coolant pump seal injection flow.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The emergency core cooling system (ECCS) analysis models the 
reactor coolant pump (RCP) seal injection flow path as a hydraulic 
flow resistance. The proposed change clarifies that RCP seal 
injection flow is a function of system conditions. The seal 
injection flow rate can vary during operation, but the hydraulic 
flow resistance is fixed by positioning the manual seal injection 
throttle valves. The resistance does not change if the valve 
adjustments are not changed. Thus, RCP seal injection flow variation 
due to changing reactor coolant system (RCS) backpressure following 
a loss of coolant accident (LOCA) is explicitly accounted for as a 
result of modeling the RCP seal injection flow path resistance.
    The proposed change does not impact the way the RCP seal 
injection flow should be established per the safety analysis and 
does not affect RCP seal integrity. The seal injection flow 
resistance only affects ECCS flow. Since ECCS flow occurs after an 
accident, the proposed change cannot impact the probability of an 
accident.
    Overall ECCS performance will remain within the bounds of the 
previously performed accident analyses since there are no hardware 
changes. The ECCS will continue to function in a manner consistent 
with the plant design basis. All design, material, and construction 
standards that were applicable prior to the proposed change are 
[still] maintained.
    The proposed change will not affect the probability of any event 
initiators. There will be no degradation in the performance of, or 
an increase in the number of challenges imposed on, safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance.
    The proposed change will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in 
the FSAR [Final Safety Analysis Report].
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The proposed change will not affect the normal method of 
plant operation. No performance requirements will be affected.
    Since the proposed change continues to assure that the assumed 
ECCS flow is available after a large break LOCA, no new accident 
scenarios, transient precursors, failure mechanisms, or limiting 
single failures are introduced as a result [of the proposed change]. 
There will be no adverse effect or challenges imposed on any safety-
related system as a result of this request.
    The proposed change does not alter the design or performance 
characteristics of the ECCS. It simply corrects the description of 
how to properly set the position of the RCP seal injection throttle 
valves in support of the ECCS flow balance assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio limits, heat flux hot 
channel factor (F<INF>Q</INF>) nuclear enthalpy rise hot channel 
factor (FN/DH), loss of coolant accident peak cladding temperature 
(LOCA PCT), peak local power density, or any other margin of safety. 
The radiological dose consequence acceptance criteria listed in the 
Standard Review Plan will continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

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

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: April 11, 2000, as supplemented by 
letters dated August 28, 2000, November 20, 2000, and April 11, 2001.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) 3.7, 3.10, and 3.22, as well as 
the Bases of TS 3.4, 3.8, 3.10, 3.19, and 3.22. The proposed changes 
would implement an alternate accident source term methodology 
previously approved by NRC. Implementation of the alternate source term 
could permit a number of plant changes that have been proposed, 
including: Permitting a slight atmospheric pressure in containment for 
a short time following a loss-of-coolant accident (LOCA), deletion of 
automatic function requirements and setpoints for containment 
particulate and gas monitors, deletion of the requirement to filter 
fuel building and containment purge exhaust during refueling, and a 
number of other related operational and configuration 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:


[[Page 34290]]


    The proposed TS changes allow relaxation of containment 
integrity requirements during refueling operations by allowing the 
personnel airlock, equipment access hatch and certain penetrations 
to remain open during fuel movement in containment. The changes also 
eliminate the requirement to filter the exhaust from containment or 
the fuel building during refueling operations. Also proposed is a 
relaxation of the current containment design basis acceptance 
criteria to allow an interval of four hours following the design 
basis LOCA until containment is depressurized to subatmospheric 
conditions. We have reviewed the proposed TS changes relative to the 
requirements of 10 CFR 50.92 and determined that a significant 
hazards consideration is not involved. Specifically, operation of 
Surry Power Station with the proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The probability remains unaffected since the accident analyses 
involve no change to a system, component or structure that affects 
initiating events for any of the accidents evaluated. The 
consequences of the reanalyzed events is expressed in terms of the 
TEDE [total effective dose equivalent] dose, which is not directly 
comparable to either the thyroid or whole body doses reported in 
existing analyses. However, even taking this comparison into 
consideration, any dose increase is not significant. Furthermore, 
the revised analysis results meet the applicable TEDE dose 
acceptance criteria for alternative source term implementation.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The implementation of the proposed changes does not create the 
possibility of an accident of a different type than was previously 
evaluated in the SAR [Safety Analysis Report]. The proposed 
Technical Specifications changes allow relaxation of these current 
requirements: (1) maintaining subatmospheric containment conditions 
following a LOCA; (2) filtration of containment & fuel building 
exhaust during fuel movement; (3) maintaining the personnel airlock, 
equipment access hatch & penetrations closed during fuel movement 
and (4) operability of containment purge isolation during refueling. 
These changes do not alter the nature of events postulated in the 
UFSAR [Updated Final SAR] nor do they introduce any unique precursor 
mechanisms. Therefore, there is no possibility for accidents of a 
different type than previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The implementation of the proposed changes does not reduce the 
margin of safety. The radiological analysis results, even though 
compared with the revised TEDE acceptance criteria, meet the 
applicable limits. These criteria have been developed for 
application to analyses performed with alternative source terms. 
These acceptance criteria have been developed for the purpose of use 
in design basis accident analyses such that meeting the stated 
limits demonstrates adequate protection of public health and safety. 
It is thus concluded that the margin of safety will not be reduced 
by the implementation of the changes.

    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: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard L. Emch.

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. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to <a href="/cdn-cgi/l/email-protection#08786c7a48667a6b266f677e"><span class="__cf_email__" data-cfemail="8cfce8fecce2feefa2ebe3fa">[email protected]</span></a>.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: November 30, 2000.
    Brief description of amendments: The amendments revise TS 5.5.13, 
``Diesel Fuel Oil Testing Program,'' to relocate the specific American 
Society for Testing and Materials (ASTM) Standard reference from the 
Administrative Controls Section of TS to a licensee-controlled 
document, i.e., the Diesel Fuel Oil Program in the Technical 
Requirements Manual (TRM). In addition, the ``clear and bright'' test 
used to establish the acceptability of new fuel oil for use prior to 
addition to storage tanks has been expanded to allow a water and 
sediment content test to be performed to establish the acceptability of 
new fuel oil in lieu of the ``clear and bright'' test.
    Date of issuance: June 13, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 122, 122, 116, and 116.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 13, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: July 31, 2000.
    Brief description of amendments: Revised Technical Specification 
(TS) Surveillance Requirement (SR) 4.5.1.d.1, concerning the 
operability of the Automatic Depressurization System, and relocated the 
existing requirements

[[Page 34291]]

in TS SR 4.5.1.d.1 and TS SR 4.5.1.d.2.c to the Technical Requirements 
Manual.
    Date of issuance: June 12, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 152 and 116.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62389).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 12, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 8, 2000, as 
supplemented on February 6, and May 7, 2001.
    Brief description of amendment: The amendment changed the technical 
specifications associated with the deletion of TS 3/4.4.1.6, ``Reactor 
Coolant Pump--Startup.''
    Date of issuance: June 13, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No: 238.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81917).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 13, 2001.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of application for amendment: January 19, 2001, as 
supplemented April 20 and May 9, 2001.
    Brief description of amendment: The amendment would change the TSs 
to extend surveillance intervals associated with the emergency diesel 
generator (EDG) engines and station batteries that are currently 
required to be completed beginning June 27, 2001. The license amendment 
would allow these requirements to be performed during the next 
refueling outage, but no later than December 31, 2001. This would 
preclude the need for a mid-cycle shutdown of the Unit.
    Date of issuance: June 11, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-74: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15926). The April 20 and May 9, 2001, supplemental letters, did not 
change the scope of the proposed action and did not change the Nuclear 
Regulatory Commission's (NRC's) preliminary no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 11, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: January 13, 2000, as 
supplemented March 7, March 30, and May 4, 2001.
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TSs) 3.6, 
``Containment'' to add Limiting Condition for Operation (LCO) and 
Allowed Outage Times (AOT) for containment isolation devices. In 
addition, the amendment provides additional information, clarification, 
and uniformity to the bases of the associated TSs.
    Date of issuance: June 8, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 155.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11061). The March 7, March 30, and May 4, 2001, letters, provided 
clarifying information that was within the scope of the original 
application, did not change the NRC staff's initial proposed no 
significant hazards consideration determination, and did not expand the 
amendment beyond the scope of the original notice (66 FR 11061).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 8, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of application for amendment: January 26, 2001, as 
supplemented by letter dated March 13, 2001.
    Brief description of amendment: The amendment changes Technical 
Specification Surveillance Requirement 3.7.9.2, ``Ultimate Heat Sink 
(UHS),'' by increasing the maximum allowable temperature of Lake 
Michigan water from 81.5  deg.F to 85  deg.F.
    Date of issuance: June 4, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 202.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13800).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 4, 2001.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: August 3, 2000, as supplemented by 
letters dated November 17, 2000, and February 14, 2001.
    Brief description of amendment: The amendment deletes Section 3.D, 
``License Term,'' from the Fort Calhoun Station, Unit No. 1 operating 
license.
    Date of issuance: June 6, 2001.
    Effective date: June 6, 2001, to be implemented within 30 days from 
the date of issuance.
    Amendment No.: 199.
    Facility Operating License No. DPR-40: The amendment revised the 
operating license.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2019).
    The November 17, 2000, and February 14, 2001, supplemental letters 
provided clarifying information, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 6, 2001.
    No significant hazards consideration comments received: No.

[[Page 34292]]

Southern California EdisonCompany, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: April 6, 2001 and supplemented 
by letter dated April 20, 2001.
    Brief description of amendments: The amendments proposed to revise 
the San Onofre Nuclear Generating Station, Units 2 and 3 Technical 
Specification Surveillance Requirements 3.8.1.2, 3.8.1.3, 3.8.1.9, 
3.8.1.10, and 3.8.1.19 to assure that an emergency diesel generator 
automatic voltage regulator (AVR) is operable and regularly tested. AVR 
operability would be demonstrated by conducting SR 3.8.1.2 and 3.8.1.3 
within the past 60 days, and any one of SR 3.8.1.9, 3.8.1.10, or 
3.8.1.19 within the past 24 months.
    Date of issuance: June 8, 2001.
    Effective date: June 8, 2001, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2-179; Unit 3-170.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22032).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: August 17, 2000, as supplemented by 
letter dated April 2, 2001. The April 2, 2001, letter requested a new 
implementation date, but did not change the August 17, 2000, 
application and the initial proposed no significant hazards 
consideration determination.
    Brief description of amendments: The amendments eliminate the need 
for the licensee to perform periodic response time testing of selected 
reactor trip system and engineered safety feature actuation system 
equipment as defined in Westinghouse report WCAP-14036-P-A, Revision 1, 
``Elimination of Periodic Protection Channel Response Time Tests.''
    Date of issuance: June 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
on Unit 1 entry in Mode 3 for Cycle 18 following the 2001 fall 
refueling.
    Amendment Nos.: 149 and 141.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2023). The supplement dated April 2, 2001, provided clarifying 
information that did not change the scope of the August 17, 2001, 
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 June 7, 2001.
    No significant hazards consideration comments received: No.

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: January 11, 2001.
    Brief description of amendments: The amendments revise TS 5.5.17, 
``Containment Leakage Rate Testing Program,'' to add an exception to 
Regulatory Guide 1.163 related to visual examination of containment 
concrete surfaces.
    Date of issuance: June 6, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 122 and 100.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22033).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 6, 2001.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant , Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: November 6, 2000.
    Description of amendment request: These amendments revised the 
Technical Specifications (TS) to allow four residual heat removal 
suppression pool cooling subsystems to be inoperable for 8 hours.
    Date of issuance: June 8, 2001.
    Effective date: June 8, 2001.
    Amendment Nos.: 241, 272, and 230.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendments revised the TS.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71139).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 3, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.3.6, ``Containment Ventilation Isolation 
Instrumentation,'' to modify the Note for Required Action B.1 such that 
it applies only to * * * Required Action and associated Completion Time 
of Condition A not met * * * This change is the result of the discovery 
of an error which occurred when the TSs were converted to the improved 
TS with issuance of License Amendment Nos. 64 and 64, for Comanche Peak 
Steam Electric Station, Units 1 and 2, on February 26, 1999.
    Date of issuance: June 4, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 86 and 86.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22034).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 4, 2001.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: September 27, 2000, as 
supplemented November 21 and December 18, 2000, and February 2, March 
2, and May 21, 2001.
    Brief description of amendment: These amendments add Technical 
Specification (TS) 3.7.14, TS 4.7.14, TS 3.7.15, TS 4.7.15, Figure 
3.7.15-1, and Figure 3.7.15-2; and revise TS 5.3.1 and TS 5.6.1.1. The 
purpose of these amendments is to increase the limit on the fuel 
enrichment from the current limit of 4.3 weight percent U<SUP>235</SUP> 
to a maximum of 4.6 weight percent U<SUP>235</SUP>, establish TS 
Limiting Conditions for Operations for the Spent Fuel Pool (SFP) boron 
concentration and fuel storage restrictions, and eliminate the value of 
uncertainties in the calculation for K<INF>eff</INF> in the SFP 
criticality calculation.

[[Page 34293]]

    Date of issuance: June 15, 2001.
    Effective date: As of the date of issuance and shall be implemented 
by December 21, 2001.
    Amendment Nos.: 227 and 208.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77929). The December 18, 2000, February 2, March 2, and May 21, 
2001, supplements contained clarifying information only, and did not 
change the initial no significant hazards consideration determination, 
or expand the scope of the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 15, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 19th day of June 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-15818 Filed 6-26-01; 8:45 am]
BILLING CODE 7590-01-P


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