[Federal Register Volume 65, Number 249 (Wednesday, December 27, 2000)]
[Notices]
[Pages 81907-81936]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-33012]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 4, 2000, through December 15, 2000. 
The last biweekly notice was published on December 13, 2000.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By January 26, 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, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and electronically from 
the ADAMS Public Library component on the NRC Web site, 
http://www.nrc.gov (the Electronic Reading Room). If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for

[[Page 81908]]

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

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

    Date of amendment request: September 29, 2000.
    Description of amendment request: The proposed amendments would 
make various changes to the Technical Specifications (TS) to support a 
change in fuel vendors from Siemens Power Corporation to General 
Electric and a transition to the use of GE 14 fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Evaluation of effect on the probability of an accident previously 
evaluated:
    1. Administrative Changes. The revisions to Current Technical 
Specifications (CTS) Sections 2.1.B, ``Thermal Power, High Pressure and 
High Flow,'' and 3.6.A, ``Recirculation Loops,'' regarding the Minimum 
Critical Power Ratio (MCPR) Safety Limit, the changes to CTS Section 
6.9.A.6.b, ``Core Operating Limits Report,'' and the changes to the 
definitions are administrative changes and will not affect the 
probability of an accident previously evaluated. These changes do not 
affect plant systems, structures, or components. No plant mitigating 
systems or functions are affected by these changes.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The changes to CTS Sections 3/4.3.C, ``Control Rod Operability,'' 3/
4.3.D, ``Maximum Scram Insertion Times,'' 3/4.3.E, ``Average Scram 
Insertion Times,'' 3/4.3.F, ``Group Scram Insertion Times,'' 3/4.3.G, 
``Control Rod Scram Accumulators,'' 3/4.3.H, ``Control Rod Coupling,'' 
and 3/4.3.I, ``Control Rod Position Indication System,'' revise the 
methodology for determining rod operability and control rod scram time 
requirements for operation. These changes do not physically alter plant 
systems, structures or components and therefore do not affect the 
probability of an accident previously evaluated.
    3. Control Rod Scram Times. The addition of required scram times 
for General Electric (GE) analyzed cores does not physically alter 
plant systems, structures or components and therefore does not affect 
the probability of an accident previously evaluated.
    4. Rod Worth Minimizer (RWM). The revision to CTS Section 3/4.3.L, 
``Rod Worth Minimizer,'' lowers the power level at which the analyzed 
rod position sequence must be followed. This change does not affect 
plant systems, structures, or components. Because there is no possible 
control rod configuration that results in a control rod worth that 
could exceed the 280 cal/gram fuel design limit, the probability of an 
accident is not increased.
    5. Transient Linear Heat Generation Rate (TLHGR). The revisions to 
CTS Section 3.11.B, ``Transient Linear Heat Generation Rate,'' add fuel 
thermal limits that are monitored to ensure that TLHGR is not violated. 
These changes do not physically alter plant systems, structures or 
components and therefore do not affect the probability of an accident 
previously evaluated.
    Evaluation of the effect on the consequences of an accident 
previously evaluated.
    1. Administrative Changes. The revisions to CTS Sections 2.1.B and 
3.6.A, regarding the MCPR Safety Limit, the changes to CTS Section 
6.9.A.6.b regarding the COLR, and the changes to

[[Page 81909]]

the definitions are administrative changes and will not affect the 
consequences of an accident previously evaluated. These changes do not 
affect plant systems, structures, or components. No plant mitigating 
systems or functions are affected by these changes.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The revisions to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I are made to ensure that appropriate scram 
times are reflected in the TS for GE methodology. The scram timing 
requirements ensure that the negative reactivity insertion rate assumed 
in the safety analyses is preserved. CTS methods ensure this by 
limiting scram times for individual rods, the average scram time, and 
local scram times (i.e., a four control rod group). The proposed 
revisions, based on the Improved Technical Specification (ITS) methods, 
ensure this by limiting the scram times for individual rods, the number 
of slow rods, and the number of adjacent slow rods. Each of these 
methods ensure equivalent protection of the assumed reactivity 
insertion rate. Therefore, there is no change to the consequences of a 
previously evaluated accident or transient.
    In addition, numerous changes to the control rod operability and 
scram timing requirements were made to reflect the ITS approach to 
these requirements. These revisions consist of administrative changes, 
more restrictive changes, and less restrictive changes. The discussion 
of each of these categories is provided below.
    Administrative changes. These consist of restructuring, 
interpretation, rearranging of requirements, and other changes not 
substantially revising an existing requirement. Therefore, these 
changes do not affect the consequences of an accident previously 
evaluated.
    More restrictive changes. These consist of changes resulting in 
added restrictions or eliminating flexibility. The more restrictive 
requirements continue to ensure that process variables, structures, 
systems and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, these changes do not involve 
an increase in the consequences of an accident previously evaluated.
    Less restrictive changes. The less restrictive changes involve 
increasing the time to complete actions, increasing the time intervals 
between required surveillances, and deleting or revising the 
applicability of certain actions. The time to complete actions and the 
surveillance frequencies are not assumed in the analysis of the 
consequences of any accidents previously evaluated, and therefore, 
cannot increase the consequences of such accidents. The deleted or 
revised actions are not assumed in the safety analyses for any 
evaluated accidents. The revised scram timing methods will result in 
operating thermal limits that will maintain the identical safety 
limits. Thus, the consequences of the evaluated accidents will not 
increase.
    3. Control Rod Scram Times. Cycle-specific analyses that use the GE 
methodology scram times will meet all of the same safety limit 
acceptance criteria. Additionally, for the non-cycle specific events in 
the Updated Final Safety Analysis Report (UFSAR), GE has determined 
that there is negligible impact on results of events which are not 
analyzed on a cycle-specific basis. Therefore, there is no change to 
the consequences of a previously evaluated accident or transient.
    4. RWM. The RWM enforces the analyzed rod position sequence to 
ensure that the initial conditions of the Control Rod Drop Accident 
(CRDA) analysis are not violated. Compliance with the analyzed rod 
position sequence, and operability of the RWM is required in Mode 1, 
``Power Operation,'' and Mode 2, ``Startup,'' when thermal power is 
less than or equal to 10% Rated Thermal Power (RTP). When thermal power 
is greater than 10% RTP, there is no possible control rod configuration 
that results in a control rod worth that could exceed the 280 cal/gm 
fuel design limit during a CRDA. Because the fuel design limit of 280 
cal/gm is not exceeded, this change to lower the Low Power Setpoint 
(LPSP) does not increase the consequences of an accident previously 
evaluated.
    5. TLHGR. The changes to this section are analytical in nature and 
do not affect plant systems, structures, or components. The changes in 
this section revise the description of fuel thermal limits that are 
monitored to ensure that the TLHGR limit is not violated. The TLHGR 
protects the fuel from 1% plastic strain and fuel centerline melt. 
Because these criteria have not changed, the consequences of an 
accident have not changed.
    Therefore, the proposed changes to the CTS do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    1. Administrative Changes. The revisions to CTS Sections 2.1.B and 
3.6.A, regarding the MCPR Safety Limit, the changes to CTS Section 
6.9.A.6.b regarding the COLR, and the changes to the definitions are 
administrative changes and will not create the possibility of a new or 
different kind of accident. These changes do not affect plant systems, 
structures, or components. No plant mitigating systems or functions are 
affected by these changes.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The changes to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I revise the control rod operability and 
scram time requirements for operation. These changes do not physically 
alter plant systems, structures or components and therefore do not 
create the possibility of a new or different kind of accident.
    3. Control Rod Scram Times. These changes do not physically alter 
plant systems, structures or components and therefore do not create the 
possibility of a new or different kind of accident.
    4. RWM. The revisions to CTS Section 3/4.3.L lower the power level 
at which the analyzed rod position sequence must be followed. This 
change does not affect plant systems, structures, or components. 
Because there is no possible control rod configuration that results in 
a control rod worth that could exceed the 280 cal/gm fuel design limit, 
no new or different type of accident is created.
    5. TLHGR. The revisions to CTS Section 3.11.B revise the 
description of fuel thermal limits that are monitored to ensure that 
TLHGR is not violated. These changes are analytical in nature and do 
not affect plant systems, structures, or components. Therefore, the 
changes do not create the possibility of a new or different kind of 
accident.
    Therefore, the proposed changes to the CTS do not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    Does the proposed change involve a significant reduction in a 
margin of safety?
    1. Administrative Changes. The revisions to CTS Sections 2.1.B and 
3.6.A, regarding the MCPR Safety Limit, the changes to CTS Section 
6.9.A.6.b, regarding the COLR, and the changes to the definitions are 
administrative changes and will not reduce the margin of safety. These 
changes do not affect plant systems, structures, or components. No 
plant mitigating systems or functions are affected by these changes.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The revisions to the CTS control rod operability and scram insertion 
times

[[Page 81910]]

ensure that the negative reactivity insertion rate assumed in the 
safety analyses is preserved. CTS methods ensure this by limiting scram 
times for individual rods, the average scram time, and the local scram 
times (i.e., a four control rod group). ITS methods ensure this by 
limiting the scram times for individual rods, the number of slow rods, 
and the number of adjacent slow rods. Each of these methods ensure 
equivalent protection of the assumed reactivity insertion rate. 
Therefore, the changes do not involve a reduction in the margin of 
safety.
    In addition, numerous changes to the control rod operability and 
scram timing requirements were made to reflect the ITS approach to 
these requirements. These revisions consist of administrative changes, 
more restrictive changes, and less restrictive changes. The discussion 
of each of these categories is provided below.
    Administrative Changes. These consist of restructuring, 
interpretation, and complex rearranging of requirements, and other 
changes not substantially revising an existing requirement. Therefore, 
these changes do not affect the margin of safety.
    More restrictive changes. These consist of changes resulting in 
added restrictions or eliminating flexibility. The more restrictive 
requirements continue to ensure that process variables, structures, 
systems and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, these changes do not reduce 
the margin of safety.
    Less restrictive changes. The less restrictive changes involve 
increasing the time to complete actions, increasing the time intervals 
between required surveillances, and deleting or revising the 
applicability of certain actions. The time to complete actions and the 
surveillance frequencies have been extended for several reasons, 
including experience showing low probability of failures, the benefit 
of allowing time to perform actions without undue haste, or due to 
compensating changes in other actions. The deleted or revised actions 
are not assumed in the safety analyses for any evaluated accidents. 
Thus, there is no significant reduction in the margin of safety.
    3. Control Rod Scram Times. The addition of required scram times 
for GE analyzed cores based on GE analysis methodology does not involve 
a reduction in the margin of safety. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR 
operating limits that will ensure the MCPR safety limit is not 
violated. Therefore, the fuel remains appropriately protected and no 
margins of safety are reduced.
    4. RWM. The RWM enforces the analyzed rod position sequence to 
ensure that the initial conditions of the CRDA analysis are not 
violated. Compliance with the analyzed rod position sequence, and 
operability of the RWM is required in Modes 1 and 2 when thermal power 
is less than or equal to 10% rated thermal power (RTP). When thermal 
power is greater than 10% RTP, there is no possible control rod 
configuration that results in a control rod worth that could exceed the 
280 cal/gm fuel design limit during a CRDA. Because the fuel design 
limit of 280 cal/gm is not exceeded above 10% RTP, this change to 
reduce the LPSP does not reduce a margin of safety.
    5. TLHGR. The addition of the ratio of Maximum Fraction of Limiting 
Power Density (MFLPD) to the Fraction of Rated Thermal Power (FRTP) 
provides thermal limit protection for GE fuel. This provides equivalent 
protection to ensure that the TLHGR limit is maintained. Therefore, the 
revisions to CTS Section 3.11.B will not reduce a margin of safety.
    Therefore, these proposed changes to the CTS do not involve a 
significant reduction in the margin safety.

Proposed Changes to ITS

    Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Evaluation of the effect on the probability of an accident 
previously evaluated.
    1. Administrative Changes. The revision to Improved Technical 
Specification (ITS) Section 5.6.5, ``Core Operating Limits Report,'' 
and the added definitions are purely administrative changes and do not 
affect the probability or consequences of an accident previously 
evaluated.
    2. Control Rod Scram Times. The revision to ITS Table 3.1.4-1, 
``Control Rod Scram Times,'' adds scram time requirements for GE 
analyzed cores. This change does not physically alter plant systems, 
structures or components and therefore does not affect the probability 
of an accident previously evaluated.
    3. Average Power Range Monitor (APRM) Gain and Setpoint. The 
revisions to ITS Section 3.2.4, ``Average Power Range Monitor (APRM) 
Gain and Setpoint,'' revise the description of fuel thermal limits that 
are monitored to ensure the TLHGR is not violated. The changes to this 
section are analytical in nature and do not affect plant systems, 
structures, or components and therefore will not affect the probability 
of an accident previously evaluated.
    Evaluation of the effect on the consequences of an accident 
previously evaluated.
    1. Administrative Changes. The revision to ITS Section 5.6.5 and 
the added definitions are purely administrative changes and do not 
affect the probability or consequences of an accident previously 
evaluated.
    2. Control rod scram times. The revisions to ITS Section 3.1.4, 
``Control Rod Scram Insertion Times,'' are made to ensure the 
appropriate scram times are reflected in the Technical Specifications 
(TS) for GE methodology. The scram timing requirements ensure that the 
negative reactivity insertion rate assumed in the safety analyses is 
preserved. Cycle specific analyses that use the GE methodology scram 
times will meet all of the same safety limit acceptance criteria. 
Additionally, for the non-cycle specific UFSAR events, GE has 
determined that there is negligible impact on the results of events 
which are not analyzed on a cycle specific basis. Therefore, there is 
no change to the consequences of a previously evaluated accident or 
transient due to the TS changes.
    3. APRM Gain and Setpoint. The revisions to ITS Section 3.2.4 will 
not increase the consequences of an accident previously evaluated. The 
changes to this section are analytical in nature and do not affect 
plant systems, structures, or components. The changes in this section 
revise the description of fuel thermal limits that are monitored to 
ensure the TLHGR limit is not violated. The TLHGR protects the fuel 
from 1% plastic strain and fuel centerline melt. Because these criteria 
have not changed, the consequences of an accident have not changed.
    Therefore, the proposed changes to the ITS do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    1. Administrative Changes. The revision to ITS Section 5.6.5 and 
the added definitions are purely administrative changes and therefore 
do not create the possibility of a new or different kind of accident.
    2. Control Rod Scram Insertion Times. The revisions to ITS Section 
3.1.4 do not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The changes to these 
sections revise the control rod scram time requirements for

[[Page 81911]]

operation. This change does not physically alter plant systems, 
structures, or components.
    3. APRM Gain and Setpoint. The revisions to ITS Section 3.2.4 will 
not create the possibility of a new or different kind of accident. The 
changes to this section are analytical in nature and do not affect 
plant systems, structures, or components. The changes in this section 
revise the description of fuel thermal limits that are monitored to 
ensure the TLHGR limit is not violated.
    Therefore, the proposed changes to the ITS do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Does the proposed change involve a significant reduction in a 
margin of safety?
    1. Administrative Changes. The revision to ITS Section 5.6.5 and 
the added definitions are purely administrative changes and do not 
affect the margin of safety.
    2. Control Rod Scram Insertion Times. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR 
operating limits that will ensure that the MCPR safety limit is not 
violated. Therefore, the fuel remains appropriately protected and no 
margins of safety are reduced.
    3. APRM Gain and Setpoint. The addition of MFLPD/FRTP provides 
thermal limit protection for GE fuel. This provides equivalent 
protection to ensure that the TLHGR limit is maintained. Therefore, the 
revisions to ITS Section 3.2.4 will not reduce a margin of safety.
    Therefore, the proposed changes to the ITS do not involve a 
significant reduction in the margin of safety.
    Based on the above evaluation, ComEd has concluded that these 
changes involve no 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: November 10, 2000.
    Description of amendment request: The proposed amendments would 
revise several sections of the Technical Specifications (TS) and add a 
new TS section to incorporate Oscillation Power Range Monitor (OPRM) 
Instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Does the change involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The proposed changes for LaSalle County Station will delete the 
thermal hydraulic instability administrative requirements and Power 
versus Flow figure and references to it from the TS, and insert a new 
TS for the OPRM instrumentation. The proposed TS will allow the 
enabling of the OPRM instrumentation trips. The deletion of the thermal 
hydraulic instability administrative requirements and Power versus Flow 
figure and the requirements to have an operable OPRM instrumentation 
trip does not have an effect on any accident previously evaluated or 
the associated accident assumptions. Thus, the proposed changes do not 
significantly increase the probability of an accident previously 
evaluated.
    The proposed changes do not adversely affect the integrity of the 
fuel cladding, reactor coolant system or secondary containment. As 
such, the radiological consequences of previously evaluated accident 
are not changed. Therefore, the proposed changes do not increase the 
consequences of an accident previously evaluated.
    Does the change create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The proposed changes do not effect the assumed accident performance 
of any structure, system, or component previously evaluated. The 
proposed changes do not introduce any new modes of system operation or 
failure mechanisms.
    The OPRM instrumentation will initiate an automatic reactor trip 
upon detection of an instability that could threaten the Minimum 
Critical Power Ratio (MCPR) safety limit. The OPRM Instrumentation 
System consists of four (4) OPRM instrumentation trip channels. When 
one OPRM instrumentation module is inoperable, the remaining redundant 
OPRM Instrumentation module in the associated OPRM trip channel 
maintains the operability of the trip channel and thus there is no loss 
of trip function redundancy.
    Thus, these proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    Boiling Water Reactors are susceptible to thermal hydraulic 
instabilities if operated at high power and low flow conditions. 10 CFR 
50, Appendix A, General Design Criterion (GDC) 10, ``Reactor design,'' 
requires the reactor core and associated coolant, control, and 
protection systems to be designed with appropriate margin to assure 
that acceptable fuel design limits are not exceeded during any 
condition of normal operation, including the effects of anticipated 
operational occurrences. Additionally, GDC 12, ``Suppression of reactor 
power oscillation,'' requires the reactor core and associated coolant, 
control, and protection systems to be designed to assure that power 
oscillations which can result in conditions exceeding acceptable fuel 
design limits are either not possible or can be reliably and readily 
detected and suppressed.
    The detection and suppression of instability is required to insure 
that the MCPR safety limit is not exceeded during a transient. The OPRM 
instrumentation will initiate an automatic reactor trip upon detection 
of an instability that could threaten the MCPR safety limit.
    The OPRM Instrumentation System consists of four (4) OPRM 
instrumentation trip channels, each trip channel consisting of two OPRM 
instrumentation modules. Each OPRM instrumentation module receives 
input from LPRMs. Each OPRM instrumentation module also receives input 
from the RPS Average Power Range Monitor (APRM) power and flow signals 
to automatically enable the trip function of the OPRM instrumentation 
module.
    Each OPRM instrumentation module is continuously tested by a self-
test function. On detection of any OPRM instrumentation module failure, 
either a ``Trouble'' or ``INOP'' alarm is activated. The OPRM 
instrumentation module provides an ``INOP'' alarm when the self-test 
feature indicates that the OPRM instrumentation module may not be 
capable of meeting its functional

[[Page 81912]]

requirements. When one OPRM instrumentation module is inoperable, the 
remaining redundant OPRM Instrumentation module in the associated OPRM 
trip channel maintains the operability of the trip channel and thus 
there is no loss of trip function redundancy. The OPRM Instrumentation 
System provides compliance with GDC 10 and GDC 12.
    The incorporation of the OPRM instrumentation into the TS will 
allow the deletion of the current thermal hydraulic instability 
administrative requirements and Power versus Flow TS Figure and 
associated actions. The OPRM instrumentation will provide the same 
level of assurance that the MCPR safety limit will not be violated for 
anticipated oscillations as that provided by the Power versus Flow TS 
Figure.
    The OPRM Instrumentation System enabled region of the Power versus 
Flow figure was adjusted to maintain the same level of protection 
against the occurrence of a thermal-hydraulic instability by 
maintaining the pre-power uprate absolute power and flow coordinates. A 
5% Power Uprate was approved for LaSalle County Station, Units 1 and 2, 
by Facility Operating License Amendments 140 and 125, respectively, in 
an NRC letter dated May 9, 2000.
    The proposed changes do not affect the margin of safety as the OPRM 
Instrumentation will initiate an automatic reactor trip upon detection 
of an instability that could threaten the MCPR safety limit.
    Thus, this proposed change does not involve a significant reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: September 29, 2000.
    Description of amendment request: The proposed amendments would 
make various changes to the Technical Specifications (TSs) to support a 
change in fuel vendors from Siemens Power Corporation to General 
Electric and a transition to the use of General Electric (GE) 14 fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

Proposed Changes to Current Technical Specifications

    Evaluation of effect on the probability of an accident previously 
evaluated:
    1. Administrative Changes. The revisions to Current Technical 
Specifications (CTS) Sections 2.1.B, ``Thermal Power, High Pressure and 
High Flow,'' and 3.6.A, ``Recirculation Loops,'' regarding the Minimum 
Critical Power Ratio (MCPR) Safety Limit, the changes to Section 3.11B, 
``Transient Linear Heat Generation Rate,'' regarding the surveillance 
to monitor Transient linear heat Generation Rate (TLHGR) using either 
the ratio of the Maximum Fraction of Limiting Power Density (MFLPD) to 
the Fraction of Rated Thermal Power (FRTP) or the Fuel Design Limiting 
Ratio for Centerline (FDLRC) Melt, and the addition of the NRC approved 
RODEX2A methodology, are administrative changes and will not affect the 
probability of an accident previously evaluated. These changes do not 
affect plant systems, structures, or components. No plant mitigating 
systems or functions are affected by these changes.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The changes to CTS Sections 3/4.3.C, ``Control Rod Operability,'' 3/
4.3.D, ``Maximum Scram Insertion Times,'' 3/4.3.E, ``Average Scram 
Insertion Times,'' 3/4.3.F, ``Group Scram Insertion Times,'' 3/4.3.G, 
``Control Rod Scram Accumulators,'' 3/4.3.H, ``Control Rod Coupling,'' 
and 3/4.3.I, ``Control Rod Position Indication System,'' revise the 
methodology for determining rod operability and control rod scram time 
requirements for operation. These changes do not physically alter plant 
systems, structures or components and therefore do not affect the 
probability of an accident previously evaluated.
    3. Control Rod Scram Times. The addition of required scram times 
for General Electric (GE) analyzed cores does not physically alter 
plant systems, structures or components and therefore does not affect 
the probability of an accident previously evaluated.
    Evaluation of the effect on the consequences of an accident 
previously evaluated.
    1. Administrative Changes. The revisions to CTS Sections 2.1.B and 
3.6.A, regarding the MCPR Safety Limit are administrative changes and 
will not affect the consequences of an accident previously evaluated. 
These changes do not affect plant systems, structures, or components. 
No plant mitigating systems or functions are affected by these changes. 
The changes to this section are analytical in nature and do not affect 
plant systems, structures, or components. The administrative changes to 
Section 3.11.B revise the description of fuel thermal limits that are 
monitored to ensure the TLHGR limit is not violated. TLHGR protects the 
fuel from 1% plastic strain and fuel centerline melt. Because these 
criteria have not changed, the consequences of an accident have not 
changed. The NRC approved burnup extension for RODEX2A has been 
demonstrated to meet all applicable design criteria. Therefore, the 
addition of the NRC approved RODEX2A methodology does not increase the 
consequences of an accident previously evaluated.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The revisions to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I are made to ensure that appropriate scram 
times are reflected in the TS for GE methodology. The scram timing 
requirements ensure that the negative reactivity insertion rate assumed 
in the safety analyses is preserved. CTS methods ensure this by 
limiting scram times for individual rods, the average scram time, and 
local scram times (i.e., a four control rod group). The proposed 
revisions, based on the Improved Technical Specification (ITS) methods, 
ensure this by limiting the scram times for individual rods, the number 
of slow rods, and the number of adjacent slow rods. Each of these 
methods ensure equivalent protection of the assumed reactivity 
insertion rate. Therefore, there is no change to the consequences of a 
previously evaluated accident or transient.
    In addition, numerous changes to the control rod operability and 
scram timing TS Sections were made to reflect the ITS approach to these 
requirements. These revisions consist of administrative changes, more 
restrictive changes, and less restrictive changes. The discussion of 
each of these categories is provided below.
    Administrative changes. These consist of restructuring, 
interpretation,

[[Page 81913]]

rearranging of requirements, and other changes not substantially 
revising an existing requirement. Therefore, these changes do not 
affect the consequences of an accident previously evaluated.
    More restrictive changes. These consist of changes resulting in 
added restrictions or eliminating flexibility. The more restrictive 
requirements continue to ensure that process variables, structures, 
systems and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, these changes do not involve 
an increase in the consequences of an accident previously evaluated.
    Less restrictive changes. The less restrictive changes involve 
increasing the time to complete actions, increasing the time intervals 
between required surveillances, and deleting or revising the 
applicability of certain actions. The time to complete actions and the 
surveillance frequencies are not assumed in the analysis of the 
consequences of any accidents previously evaluated, and therefore, 
cannot increase the consequences of such accidents. The deleted or 
revised actions are not assumed in the safety analyses for any 
evaluated accidents. The revised scram timing methods will result in 
operating thermal limits that will maintain the identical safety 
limits. Thus, the consequences of the evaluated accidents will not 
increase.
    3. Control Rod Scram Times. Cycle-specific analyses that use the GE 
methodology scram times will meet all of the same safety limit 
acceptance criteria. Additionally, for the non-cycle specific events in 
the Updated Final Safety Analysis Report (UFSAR), GE has determined 
that there is negligible impact on results of events which are not 
analyzed on a cycle-specific basis. Therefore, there is no change to 
the consequences of a previously evaluated accident or transient.
    Therefore, the proposed changes to the CTS do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    1. Administrative Changes. The revisions to CTS Sections 2.1.B and 
3.6.A, regarding the MCPR Safety Limit, the revisions to CTS Section 
3.11.B to revise the description of TLHGR, and the addition of the NRC 
approved RODEX2A methodology are administrative changes and will not 
create the possibility of a new or different kind of accident. These 
changes do not affect plant systems, structures, or components. No 
plant mitigating systems or functions are affected by these changes.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The changes to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I revise the control rod operability and 
scram time requirements for operation. These changes do not physically 
alter plant systems, structures or components and therefore do not 
create the possibility of a new or different kind of accident.
    3. Control Rod Scram Times. These changes do not physically alter 
plant systems, structures or components and therefore do not create the 
possibility of a new or different kind of accident.
    Therefore, the proposed changes to the CTS do not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    Does the proposed change involve a significant reduction in a 
margin of safety?
    1. Administrative Changes. The revisions to CTS Sections 2.1.B and 
3.6.A, regarding the MCPR Safety Limit, and the changes to CTS Section 
3.11.B regarding the surveillance to monitor TLHGR, and the addition of 
the NRC approved RODEX2A methodology are administrative changes and 
will not reduce the margin of safety. These changes do not affect plant 
systems, structures, or components. No plant mitigating systems or 
functions are affected by these changes.
    2. Control Rod Operability and Scram Insertion Times Methodology. 
The revisions to the CTS control rod operability and scram insertion 
times ensure that the negative reactivity insertion rate assumed in the 
safety analyses is preserved. CTS methods ensure this by limiting scram 
times for individual rods, the average scram time, and local scram 
times (i.e., a four control rod group). ITS methods ensure this by 
limiting the scram times for individual rods, the number of slow rods, 
and the number of adjacent slow rods. Each of these methods ensure 
equivalent protection of the assumed reactivity insertion rate. 
Therefore, the changes do not involve a reduction in the margin of 
safety.
    In addition, numerous changes to the control rod operability and 
scram timing TS Sections were made to reflect the ITS approach to these 
requirements. These revisions consist of administrative changes, more 
restrictive changes, and less restrictive changes. The discussion of 
each of these categories is provided below.
    Administrative Changes. These consist of restructuring, 
interpretation, and complex rearranging of requirements, and other 
changes not substantially revising an existing requirement. Therefore, 
these changes do not affect the margin of safety.
    More restrictive changes. These consist of changes resulting in 
added restrictions or eliminating flexibility. The more restrictive 
requirements continue to ensure that process variables, structures, 
systems and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, these changes do not reduce 
the margin of safety.
    Less restrictive changes. The less restrictive changes involve 
increasing the time to compete actions, increasing the time intervals 
between required surveillances, and deleting or revising the 
applicability of certain actions. The time to complete actions and the 
surveillance frequencies have been extended for several reasons, 
including experience showing low probability of failures, the benefit 
of allowing time to perform actions without undue haste, or due to 
compensating changes in other actions. The deleted or revised actions 
are not assumed in the safety analyses for any evaluated accidents. 
Thus, there is no significant reduction in the margin of safety.
    3. Control Rod Scram Times. The addition of required scram times 
for GE analyzed cores based on GE analysis methodology does not involve 
a reduction in the margin of safety. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR 
operating limits that will ensure the MCPR safety limit is not 
violated. Therefore, the fuel remains appropriately protected and no 
margins of safety are reduced.
    Therefore, these proposed changes to the CTS do not involve a 
significant reduction in the margin safety.

Proposed Changes to ITS

    Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Evaluation of the effect on the probability of an accident 
previously evaluated.
    1. Administrative change. The addition of the NRC approved RODEX2A 
methodology is an administrative change and will not affect the 
probability of an accident previously evaluated. This change does not 
affect plant systems, structures, or components. No plant mitigating 
systems or functions are affected by these changes.
    2. Control Rod Scram Times. The revision to ITS Table 3.1.4-1, 
``Control

[[Page 81914]]

Rod Scram Times,'' adds scram time requirements for GE analyzed cores. 
This change does not physically alter plant systems, structures or 
components and therefore does not affect the probability of an accident 
previously evaluated.
    Evaluation of the effect on the consequences of an accident 
previously evaluated.
    1. Administrative Change. The NRC approved burnup extension for 
RODEX2A has been demonstrated to meet all applicable design criteria. 
Therefore, the addition of the NRC approved RODEX2A methodology does 
not increase the consequences of an accident previously evaluated.
    2. Control Rod Scram Times. The revisions to ITS Section 3.1.4, 
``Control Rod Scram Insertion Times,'' are made to ensure the 
appropriate scram times are reflected in the Technical Specifications 
(TS) for GE methodology. The scram timing requirements ensure that the 
negative reactivity insertion rate assumed in the safety analyses is 
preserved. Cycle specific analyses that use the GE methodology scram 
times will meet all of the same safety limit acceptance criteria. 
Additionally, for the non-cycle specific events in the UFSAR, GE has 
determined that there is negligible impact on the results of events 
which are not analyzed on a cycle specific basis. Therefore, there is 
no change to the consequences of a previously evaluated accident or 
transient due to the TS changes.
    Therefore, the proposed changes to the ITS do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    1. Administrative Change. The addition of the NRC approved RODEX2A 
methodology is an administrative change and will not create the 
possibility of a new or different kind of accident. This change does 
not affect plant systems, structures, or components. No plant 
mitigating systems or functions are affected by this change.
    2. Control Rod Scram Insertion Times. The revisions to ITS Section 
3.1.4 do not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The changes to these 
sections revise the control rod scram time requirements for operation. 
This changes does not physically alter plant systems, structures, or 
components.
    Therefore, the proposed changes to the ITS do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Does the proposed change involve a significant reduction in a 
margin of safety?
    1. Administrative Change. The addition of the NRC approved RODEX2A 
methodology is an administrative change and will not reduce the margin 
of safety. This change does not affect plant systems, structures, or 
components. No plant mitigating systems or functions are affected by 
this change.
    2. Control Rod Scram Insertion Times. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR 
operating limits that will ensure that MCPR safety limit is not 
violated. Therefore, the fuel remains appropriately protected and no 
margins of safety are reduced.
    Therefore, the proposed changes to the ITS do not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: November 21, 2000.
    Description of amendment request: The proposed amendment would 
approve a proposed change to the licensing basis regarding the timing 
of the release of fission products following an accident. The proposed 
change is based upon one of the insights established in NUREG-1465, 
``Accident Source Terms for Light Water Nuclear Power Plants,'' which 
recognizes that there is a delay in the release of fission products 
from the reactor fuel following a postulated design-basis loss-of-
coolant accident (LOCA). The timing of fission product release from 
perforated fuel rods (i.e., the gap activity release) is based on the 
boiling-water reactor (BWR)-specific value of the timing of the gap 
activity release phase of a LOCA as calculated in the BWR Owners Group 
(BWROG) Report, ``Prediction of the Onset of Fission Gas Release From 
Fuel in Generic BWR,'' NEDC-32963A, dated March 2000, as previously 
approved by the NRC staff. This BWROG report would be added (as 
Reference 4) to the list of references in Updated Final Safety Analysis 
Report (UFSAR) Section 15.6.7. The licensing basis change to UFSAR 
Section 15.6.5.5.1, ``Fission Product Release From Fuel,'' would add 
the following: ``For primary containment isolation purposes, the 
activity from the damaged core is assumed to be released into the 
containment at 121 seconds following the accident. This timing 
assumption recognizes conclusions derived from the source term studies 
described in NUREG-1465, Regulatory Guide 1.183 and Reference 4. * * * 
The results of this Table [15.6.5-2, which presents the airborne 
activity in the containment] conservatively assume activity released 
from the core enters the drywell at accident time zero.'' UFSAR Section 
15.6.5.5.2, ``Fission Product Transport to the Environment,'' would be 
similarly supplemented to state, ``The results in this Table [15.6.5-3, 
which gives the fission product release to the environment due to 
containment leakage and leakage from engineered safety feature 
components outside containment] conservatively assume activity released 
from the core enters the drywell at accident time zero.'' UFSAR Section 
15.6.5.5.3, ``Results,'' would be supplemented to state, ``Dose 
associated with coolant activity release in the first 121 seconds of 
the accident is not included in this Table [15.6.5-4, which presents 
the calculated exposures for the design basis analysis]. Its 
contribution to the accident dose is insignificant (on the order of 2 
rem [to the] thyroid at the Exclusion Area Boundary).''
    The effect of the NRC staff's approval of the proposed amendment is 
to allow the licensee, in accordance with 10 CFR 50.59, to increase the 
automatic closure times for selected primary containment isolation 
valves (PCIVs) (i.e., those PCIVs credited for limiting post-accident 
doses to both control room personnel and to offsite individuals). 
Valves with closure times based on other requirements (i.e., system 
performance requirements, equipment qualification, high-energy line 
break mitigation, or other regulatory requirements) would not be 
affected by the proposed change.
    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

[[Page 81915]]

consideration, which is presented below:
    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change takes credit for one of the alternative source 
term (AST) insights contained in NUREG-1465 which recognizes that 
fission product release from a fuel assembly is not instantaneous in a 
design basis accident. Implementation of this change into the licensing 
basis will be used to justify an increase in the maximum allowable 
closure times for primary containment isolation valves. A change in the 
timing of the gap release does not affect the precursors for any 
accident or transient previously evaluated as part of the Fermi 2 
licensing basis. Therefore, there is no increase in the probability of 
any accident.
    A plant specific radiological analysis has been performed to 
evaluate the effects of extending the maximum allowable valve closure 
times on accident dose consequences. This evaluation utilized the 
insights contained in NUREG-1465 * * * and NEDC-32963A * * * to justify 
no gap activity release during the initial 121 seconds of the accident. 
Therefore, during this period, the only releases are from reactor 
coolant activity. Assuming the maximum coolant iodine activity 
permitted in the Technical Specifications, the 2-hour Exclusion Area 
Boundary (EAB) dose associated with this release has been 
conservatively estimated to be less than 2 rem thyroid. This dose 
represents a small fraction of the LOCA dose evaluated in the UFSAR and 
is significantly lower than the 300 rem thyroid dose acceptance limit 
in 10 CFR Part 100.
    UFSAR Figures 6.2-9 and 6.2-11 show the DBA [design-basis accident] 
LOCA primary containment pressure response. These figures indicate that 
drywell pressure peaks at around 5 seconds into the accident before 
gradually dropping off; therefore, PCIVs would not be required to close 
against increased containment pressure as a result of this change.
    Utilizing all of the insights contained in NUREG-1465, would result 
in a reduction in the calculated dose. However, because this request is 
for a selective implementation of the AST scope, crediting only the 
timing of the gap activity release, the long term dose calculations 
based on TID-14844 in the UFSAR are not changed. Therefore, it is 
concluded that the proposed change does not significantly increase the 
consequences of a previously evaluated accident.
    2. The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The primary containment isolation system is designed to prevent the 
unfiltered release of radioactive material to the environs following an 
accident. Therefore, the system is relied upon to mitigate the dose 
consequences of an accident. The proposed change recognizes the time 
delay before fission products are released into the containment as a 
result of fuel damage and allows for the adjustment of the maximum PCIV 
closure times accordingly. This change does not affect the function of 
the primary containment isolation system. The relaxation in valve 
closure times will be applied only to valves that do not have other 
system performance requirements on isolation time. Therefore, the 
proposed change does not create the potential for a new or different 
kind of accident from any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed change revises the Fermi 2 licensing basis for the 
offsite dose calculations during the initial 121 seconds of a LOCA 
scenario. For this period of time, only coolant activity release is 
postulated. No fission product release from perforated fuel rods is 
assumed. All other assumptions, bases and methodologies used in the 
long-term offsite dose calculations remain unchanged. The total dose 
shown in UFSAR Table 15.6.5-4 does not significantly increase due to 
the delay in the fission product release. The total amount of 
radioactivity remains the same and is bounded by the limits established 
in 10 CFR 100. The dose associated with coolant activity release in the 
initial 121 seconds of the accident has been determined to be 
insignificant. Therefore, the proposed change will not result in a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: Claudia M. Craig.

Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: November 22, 2000.
    Description of amendment request: The proposed amendment would 
change the pressure-temperature limit curves of Figures 3.6.1, 3.6.2, 
and 3.6.3 of Pilgrim's Technical Specifications (TSs) to cover 
operation between 20, 32, and 48 Effective Full Power Years. Also 
changes to the Bases section consistent with the TS changes are 
proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:
    The proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The licensee has proposed to adopt a change in the calculation 
methodology for the pressure-temperature limits based upon Code Cases 
N-640 and N-588. The code cases were developed using knowledge gained 
through years of industry experience. Pressure-temperature curves 
developed using the allowances of Code Cases N-640 and N-588 yield more 
operating margin. However, the experience gained in the areas of 
fracture toughness of materials and pre-existing undetected defects 
show that some of the previous assumptions used for the calculation of 
pressure-temperature limits are overly conservative. There are no 
physical changes to the plant being introduced by the proposed changes 
to the pressure-temperature curves. The proposed changes do not modify 
the reactor coolant pressure boundary, (i.e., there are no changes in 
operating pressure, materials or seismic loading). The proposed changes 
do not adversely affect the integrity of the reactor coolant pressure 
boundary such that its function in the control of radiological 
consequences is affected. Therefore, providing the allowances of the 
subject code cases in developing the pressure-temperature limit curves 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluate. The proposed changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes represent a change in the methodology in how 
the

[[Page 81916]]

pressure-temperature curves were generated. The proposed changes 
provide more operating margin in the pressure-temperature limit curves 
for in-service leakage and hydrostatic pressure testing, non-nuclear 
heatup and cooldown, and criticality. However, compliance with the 
proposed pressure-temperature curves will ensure conditions in which 
brittle fracture of primary coolant pressure boundary materials is 
possible will be avoided because such compliance with the proposed 
pressure-temperature curves provides sufficient protection against a 
non-ductile-type fracture of the reactor pressure vessel. Therefore, no 
new modes of operation are introduced nor will the changes create any 
failure mode not bounded by the previously evaluated accidents. 
Further, the proposed changes to the pressure-temperature curves do not 
affect any activities or equipment and are not assumed in any safety 
analysis to initiate any accident sequence. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not involve a significant reduction in a 
margin of safety.
    The proposed changes reflect an update of the pressure-temperature 
curves. The revised curves are based on the latest U.S. Nuclear 
Regulatory Commission and American Society of Mechanical Engineers 
(ASME) guidance. The revised pressure-temperature limits have been 
developed using the Kic fracture toughness curve shown in 
the ASME Boiler and Pressure Vessel (B&PV) Code Section XI, Appendix A, 
Figure A-2000-1, in lieu of the KIa fracture toughness curve 
shown in ASME B&PV Code Section XI, Appendix G, Figure G-2010-1, as the 
lower bound fracture toughness. The other margins involved with the 
ASME B&PV Code, Section XI, Appendix G process of determining pressure-
temperature limit curves remain unchanged.
    These revised pressure-temperature limits, although less 
restrictive than the current limits, are established in accordance with 
current regulations and the latest ASME Code information. The revised 
pressure-temperature curves provide more operating margin and, thus, 
more operational flexibility than the current pressure-temperature 
curves. However, industry experience since the inception of the 
pressure-temperature limits in 1974 confirms that some of the original 
methodologies used to develop pressure-temperature curves are overly 
conservative. Accordingly, ASME Code Cases N-640 and N-588 take 
advantage of the acquired knowledge by establishing more realistic 
methodologies for the development of pressure-temperature curves. 
Therefore, operational flexibility is gained and an acceptable margin 
of safety to reactor pressure vessel non-ductile type fracture is 
maintained. No plant safety limits, setpoints, or design parameters are 
adversely affected by the proposed changes. Therefore, the proposed 
changes do not involve a significant reduction in a margin of safety.
    Based on the staff's analysis, 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: November 30, 2000.
    Description of amendment request: The proposed amendment would 
relocate the boration systems requirements from the Technical 
Specifications (TSs) to the Technical Requirements Manual (TRM).
    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 boration systems, BAMT [boric acid makeup tank], Boric Acid 
Makeup Pumps, and Charging Pumps, are part of the CVCS [chemical and 
volume and control system], which functions to maintain Reactor Coolant 
System inventory and chemistry. The boration system functions will 
continue to be maintained in accordance with their associated design 
requirements. During accident conditions when a boration source is 
required for accident mitigation, the RWT [refueling water tank] 
provides suction for the High Pressure Safety Injection (HPSI) and Low 
Pressure Safety Injection (LPSI) pumps. The CVCS boration systems are 
not credited in the mitigation of any accidents. Therefore, the dose 
consequences associated with accident analysis will be unchanged. The 
HPSI, LPSI pumps and RWT are required by Technical Specifications.
    Based on an evaluation of the criterion listed in 10 CFR 
50.36(c)(2)(ii), the relocation of the CVCS boration systems to the TRM 
is acceptable. No changes will be made to these systems that will 
affect their current operation.
    Therefore, this change does not involve a significant increase in 
the probability of [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 design and functions of the Boric Acid Makeup Tanks, Boric Acid 
Makeup Pumps, Charging Pumps and associated flow paths will continue to 
be maintained. These systems are not accident initiators. Because the 
proposed amendment will not change the design, configuration or method 
of operation of the plant, it will not create the possibility of a new 
or different kind of accident.
    Safety Analysis Report (SAR) Chapter 15 provides the analysis of 
accidents that are considered credible. The Uncontrolled Control 
Element Assemblies (CEA) withdrawal from a subcritical or a critical 
condition, Boration Dilution Event, and Loss of Coolant Accident (LOCA) 
were evaluated in relationship to relocating these specifications to 
the TRM. Boric acid injection via the CVCS system was not credited in 
mitigating any of these 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 movement of these TSs to the TRM does not reduce the existing 
TSs or surveillance requirements. The proposed change does not change 
the design function for any of these components. Additionally, none of 
the boration systems contained in these specifications are credited in 
any accident analysis. The systems are used to maintain RCS [reactor 
coolant system] chemistry and inventory and this function will be 
maintained.
    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

[[Page 81917]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 10, 2000.
    Description of amendment request: Entergy Operations, Inc. is 
proposing that the Grand Gulf Nuclear Station (GGNS) Operating License 
be amended to modify those Technical Specifications (TS) required to 
support GGNS Cycle 12 operation. The modifications would include a 
change to the Safety Limit Minimum Critical Power Ratio (SLMCPR) 
reported in TS 2.1.1.2, and the references for analytical methods used 
to determine reactor core operating limits listed in TS 5.6.5. 
Specifically, the proposed amendment reflects a decrease of the two 
recirculation loop SLMCPR limit from 1.09 to 1.08, with the single 
recirculation loop SLMCPR limit remaining unchanged at 1.10. The 
proposed changes are necessary in order to reflect the Nuclear 
Regulatory Commission (NRC) approved methods used in determining the 
GGNS Cycle 12 core operating limits, and reflect the safety limit 
changes for the Cycle 12 mixed core consisting of Siemens Power 
Corporation (SPC) ATRIUM-10 reload fuel and General Electric (GE) GE-11 
reactor fuel.
    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 2.1.1 as that limit which 
``ensures that during normal operation and during Anticipated 
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in the 
core do not experience transition boiling.'' The MCPR safety limit 
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-approved 
methodologies. The analyses for GGNS Cycle 12 have concluded that a 
two-loop MCPR safety limit of 1.08, based on the application of Siemens 
Power Corporation's 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 (unchanged), 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 occurrences (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 SPC methods in 
the list of NRC approved methods applicable to GGNS. 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 Ratio (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 12 is of a design compatible 
with the co-resident GE-11. Therefore, the introduction of ATRIUM-10 
fuel into the Cycle 12 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 
12 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 
Siemens Power Corporation'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 Siemens Power Corporation 
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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 8, 2000.
    Description of amendment request: The proposed amendment will 
delete Technical Specification (TS) 3/4.4.1.6, ``Reactor Coolant Pump-
Startup,'' from the Beaver Valley Power Station (BVPS) TSs. This is 
accompanied by moving the secondary side water temperature to

[[Page 81918]]

cold leg temperature difference Reactor Coolant Pump (RCP) start 
requirement to existing Reactor Coolant System (RCS) TSs and deleting 
the pressurizer level requirement from Unit 1 TS 3/4.4.1.6. Unit 2 TS 
3/4.4.1.6 does not contain the pressurizer level requirement. The RCS 
TSs affected are TS 3/4.4.1.2, ``Reactor Coolant System--Hot Standby,'' 
(for Unit 2 only) and 3/4.4.1.3, ``Reactor Coolant System--Shutdown,'' 
(both units).
    Changes to the affected Bases of the Technical Specifications will 
also be made.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes will not significantly increase the 
probability of an accident previously evaluated in the BVPS Updated 
Final Safety Analysis Report (UFSAR) because accident initiation 
probabilities are independent of these changes. The proposed changes do 
not adversely affect any accident initiating events. The assumptions of 
the safety analysis are not changed by this license amendment request. 
The applicable concern associated is the possibility of 
overpressurizing the Reactor Coolant System (RCS) when a Reactor 
Coolant Pump (RCP) is started in a non-isolated loop. Adhering to a 
maximum secondary to primary side temperature difference (Technical 
Specifications 3/4.4.1.2, Reactor Coolant System--Hot Standby, Unit 2 
only, and 3/4.4.1.3, Reactor Coolant System--Shutdown, both units), 
before an RCP is started and the operability of the OPPS (Technical 
Specification 3/4.4.9.3, Overpressure Protection Systems, for both 
units), which uses the PORVs as a pressure relief device, prevents 
this. The existing Technical Specifications specify when the OPPS is to 
be operable, the maximum secondary to primary side temperature 
difference permitted, and the operability requirements imposed on the 
PORVs.
    The consequences associated with the starting of an RCP and 
potential overpressurization of the RCS also are not changed by the 
proposed license amendment. None of the accident prevention or 
mitigation controls or capabilities have been changed. Reactor Coolant 
Pump start restrictions are retained with the Technical Specifications, 
except for the pressurizer level requirement for BVPS Unit 1. This 
requirement has been shown to be unnecessary in preventing RCS 
overpressurization because the analysis assumes a water solid 
pressurizer when at least one PORV is operable. The safety analysis has 
shown that the temperature difference requirement is sufficient to 
preclude RCS overpressurization provided one PORV is available for 
pressure relief. As a result, the proposed changes will not affect any 
accident analysis consequences.
    The Technical Specifications continue to specify the maximum 
secondary to primary side temperature difference, when the OPPS is to 
be enabled, and the operability requirements for the PORVs. These 
requirements are not altered by this license amendment request and will 
continue to assure that the OPPS analysis assumptions are met. It is 
sufficient to specify the temperature difference restriction for only 
Unit 2 Technical Specification 3/4.4.1.2 because the Unit 1 OPPS 
enabling temperature is not within the applicability of Technical 
Specification 3/4.4.1.2; i.e., Mode 3, whereas the OPPS enabling 
temperature is for Unit 2. Therefore, assurance is provided that the 10 
CFR 50 Appendix G limits are not exceeded and that this proposed change 
is acceptable.
    The Bases and editorial changes, needed to meet format requirements 
and reflect the deletion of Technical Specification 3/4.4.1.6, have no 
effect on accident probabilities or consequences.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not modify the manner in which any plant 
equipment is maintained. The equipment used to prevent RCS 
overpressurization is not altered by the proposed changes. 
Specification of the number of PORVs required to be operable when the 
OPPS is enabled, and at what temperature the OPPS is required, will 
continue to be retained in Technical Specification 3/4.4.9.3, 
Overpressure Protection Systems. The necessary RCP start restrictions 
assumed in the safety analysis are not affected by the proposed 
changes. It has been shown that deleting the pressurizer level 
requirement for Unit 1 is consistent with the OPPS analysis. To assure 
the 10 CFR 50 Appendix G limits are not violated, the necessary 
requirements for starting an RCP in a non-isolated loop are retained 
within the Technical Specifications. Therefore, the analysis of an 
overpressurization of the RCS due to a heat input transient caused by 
starting an idle RCP is not changed by this license amendment request.
    The Bases and editorial changes, needed to meet format requirements 
and reflect the deletion of Technical Specification 3/4.4.1.6, will not 
affect the creation of accidents. The OPPS analysis has demonstrated 
that an RCP can be started with a water solid RCS, provided the 
secondary to primary side temperature difference requirement is met, 
and a single PORV is available for pressure relief, without violating 
10 CFR 50 Appendix G limits.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated for BVPS.
    3. Does the change involve a significant reduction in a margin of 
safety?
    The margin of safety associated with starting an RCP in a non-
isolated loop is the ability of a single OPPS PORV to relieve the 
potential RCS pressure increase without violating 10 CFR 50 Appendix G 
limits. This is maintained by meeting the secondary side water 
temperature to cold leg temperature difference and PORV operability 
requirements imposed by the Technical Specifications. These Technical 
Specification requirements are not altered by the proposed changes. The 
only deletion being proposed is the elimination of the pressurizer 
level requirement for BVPS Unit 1. This requirement has been shown to 
be unnecessary in meeting 10 CFR 50 Appendix G limits because the OPPS 
analysis assumes a water solid pressurizer and at least one OPPS PORV 
is operable. Starting an RCP with both OPPS PORVs not operable is not 
consistent with the RCS venting actions required by Technical 
Specification 3/4.4.9.3. In order to comply with the venting required 
actions with neither PORV operable, the RCS must be depressurized or in 
the process of being depressurized. Depressurization of the RCS would 
preclude starting an RCP. In order to start an RCP, the RCS must be 
pressurized to ensure a minimum pressure differential exists across the 
No. 1 seal of the RCP. Therefore, the PORV related requirements of Unit 
1 Technical Specification 3/4.4.1.6 are sufficiently addressed by 
Technical Specification 3/4.4.9.3. By eliminating PORV operability 
requirements from Unit 1 Technical Specification 3/4.4.1.6,

[[Page 81919]]

the Technical Specifications become more consistent between the two 
units and with the Standard Technical Specifications. All other RCP 
start and OPPS requirements are retained within the Technical 
Specifications and associated Bases sections.
    The Bases and editorial changes, needed to meet format requirements 
and reflect the deletion of Technical Specification 3/4.4.1.6, will not 
affect the margin of safety. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety regarding meeting 
10 CFR 50 Appendix G limits.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: November 9, 2000.
    Description of amendment request: The proposed changes would revise 
the action statements of the Davis-Besse Nuclear Power Station 
Technical Specifications (DBNPS) (TS) Limiting Condition for Operation 
(LCO) 3.5.2 and 3.6.2.1. This proposal would extend the allowed outage 
time for one Low Pressure Injection (LPI) System/Decay Heat Cooler 
train of an Emergency Core Cooling System (ECCS) subsystem from 72 
hours to 7 days (168 hours) for LCO 3.5.2. One Containment Spray System 
train may be impacted by the inoperability of the associated LPI train. 
Therefore, an extension of the allowed outage time for one train of the 
Containment Spray System from 72 hours to 7 days for LCO 3.6.2.1 is 
also being proposed, as well as new information to be added to TS Bases 
Section 3/4.5.2 and 3/4.5.3 to clarify the TS LCO 3.5.2 requirements. 
These proposed changes are based on the Babcock & Wilcox Owners Group 
(BWOG) Topical report BAW-2295A, Revisions 1 & 2, ``Justification for 
the Extension of Allowed Outage Time for Low Pressure Injection and 
Reactor Building Spray System,'' dated October 9, 1998.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit No. 1, in accordance with these changes 
would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because, as demonstrated in the Babcock & 
Wilcox Owners Group's Topical Report BAW-2295A, Revisions 1 and 2, 
Justification for Extension of Allowed Outage Time for Low Pressure 
Injection and Reactor Building Spray Systems, no accident initiators, 
conditions, or assumptions are affected by the proposed changes to 
extend the allowed outage time (AOT) from 72 hours to 7 days for one 
inoperable train of Low Pressure Injection (LPI) in Technical 
Specification (TS) 3/4.5.2 Emergency Core Cooling Systems--ECCS 
subsystems--Tavg  280 deg.F or Containment Spray 
in TS 3/4.6.2.1, Containment Systems--Depressurization and Cooling 
Systems--Containment Spray System. The proposed change to TS Bases 
Section 3/4.5.2 and 3/4.5.3 are discussions of the present TS Limiting 
Condition for Operation (LCO) which do not affect the probability of an 
accident.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because an extension in the allowable 
outage time from 72 hours to 7 days for one inoperable train will not 
affect any previously evaluated accidents. The proposed changes to the 
TS Bases discuss the present TS LCO and do not affect the consequences 
of an accident. The proposed changes do not alter the source term, 
containment isolation, or allowable radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new failure 
mode or transient is introduced since the proposed changes do not 
involve a plant modification or allow operation of any plant systems, 
structures, or components in a manner not addressed in the DBNPS Design 
Basis Accident analyses.
    3. Not involve a significant reduction in a margin of safety 
because extending the allowed outage time to 7 days for one inoperable 
train does not impact any assumptions or inputs in the DBNPS Updated 
Safety Analysis Report. The proposed changes have been evaluated and 
determined that the extended allowed outage time is consistent with 
safe operation considering the redundant systems of required features 
and the administrative controls in place for removing this equipment 
from service. The proposed TS Bases changes reflect the existing TS LCO 
and, therefore, do not reduce a margin of safety.
    On the basis of the above, the DBNPS 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.

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

    Date of amendment request: November 9, 2000.
    Description of amendment request: The proposed change would 
relocate Technical Specification (TS) 3/4.4.9.2 to the Davis Besse 
Nuclear Power Station (DBNPS) Updated Safety Analysis Report (USAR) 
Technical Requirements Manual (TRM). A corresponding change to the TS 
index is also proposed. Relocation of TS 3/4.4.9.2 to the USAR TRM will 
allow future proposed changes to the requirements to be evaluated in 
accordance with 10 CFR 50.59 and implemented if prior Nuclear 
Regulatory Commission (NRC) approval is not required. The proposed 
change is in accordance with the requirements of 10 CFR 50.36 and the 
relocation guidance provided in the NRC's ``Final Policy Statement on 
TS Improvements for Nuclear Reactors,'' dated July 22, 1993. The 
proposed change is also in accordance with the guidance provided by the 
improved ``Standard Technical

[[Page 81920]]

Specifications--Babcock & Wilcox Plants,'' NUREG-1430, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit No. 1, in accordance with these changes 
would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. No previously analyzed accident scenario is 
changed, and initiating conditions and assumptions remain as previously 
analyzed.
    The proposed change would relocate TS 3/4.4.9.2 ``Reactor Coolant 
System--Pressurizer,'' to the DBNPS Updated Safety Analysis Report 
(USAR) Technical Requirements Manual (TRM). TS 3/4.4.9.2 provides 
temperature limits for the Pressurizer based on its fatigue analysis 
design criteria. The proposed change to remove this TS is in accordance 
with 10 CFR 50.36 and the NRC's ``Final Policy Statement on TS 
Improvements for Nuclear Power Reactors,'' dated July 22, 1993. The 
proposed change is also consistent with the improved ``Standard 
Technical Specifications--Babcock and Wilcox Plants,'' NUREG-1430, 
Revision 1. A corresponding change to the TS Index page V that removes 
reference to the Pressurizer Pressure/Temperature Limits is an 
administrative change.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed change does not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed change does not 
alter the source term, containment isolation or allowable radiological 
releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new failure 
mode is introduced since the proposed relocation does not involve a 
modification or change in operation of any plant systems, structures, 
or components. No new, or different types of failures or accident 
initiators are introduced by the proposed change.
    3. Not involve a significant reduction in a margin of safety 
because the proposed change is administrative in nature, consisting of 
the relocation of certain TS requirements into a licensee-controlled 
document, and has no bearing on the margin of safety which exists in 
the present TS or Updated Safety Analysis Report (USAR).
    On the basis of the above, the DBNPS 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 No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of amendment request: October 30, 2000.
    Description of amendment request: The proposed amendment would 
revise the St. Lucie Unit 1 Technical Specification (TS) 3.9.4, 
Containment Penetrations. TS 3.9.4 requires a personnel airlock (PAL) 
door to be closed during core alterations or movement of irradiated 
fuel within containment. The proposed change would allow both 
containment PAL doors to be open during core alterations and movement 
of irradiated fuel in containment provided: (a) that at least one 
personnel airlock door is capable of being closed; (b) the plant is in 
MODE 6 with at least 23 feet of water above the fuel; and (c) a 
designated individual is available outside the PAL to close the door. 
Operability of the containment PAL door includes the requirements that 
the door is capable of being closed and that any cables or hoses across 
the PAL door have quick-disconnects to ensure the door is capable of 
being closed in a timely manner.
    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 change to TS 3.9.4 would allow the containment 
personnel airlock (PAL) doors to be open during fuel movement or core 
alterations. Currently, a single PAL door is closed during fuel 
movement or core alterations to prevent the escape of radioactive 
material in the event of an in-containment fuel handling accident. The 
PAL is not an initiator of an accident. Whether the PAL doors are open 
or closed during fuel movement and core alterations has no affect on 
the probability of any accident previously evaluated.
    Allowing the PAL doors to be open during fuel movement or core 
alterations does not significantly increase the consequences from a 
fuel handling accident. The calculated offsite doses are well within 
the limits of 10 CFR Part 100. In addition, the calculated doses are 
larger than the expected doses because the calculation does not 
incorporate the closing of the PAL doors after the containment is 
evacuated. The proposed change should significantly reduce the dose to 
workers in containment in the event of a fuel handling accident by 
reducing the time required to evacuate the containment.
    The changes being proposed do not affect assumptions contained in 
plant safety analyses or the physical design of the plant, nor do they 
affect other Technical Specifications that preserve safety analysis 
assumptions. Therefore, 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 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 change to Technical Specification 3.9.4, Containment 
Penetrations, affects a previously evaluated fuel handling accident. 
Both the current and the reanalyzed fuel handling accident analysis 
assume that all of the iodine and noble gases that become airborne 
within the containment escape and reach the site boundary and low 
population zone with no credit taken for filtration, the containment 
building barrier, or for decay or deposition taken. Since the

[[Page 81921]]

proposed change does not involve the addition or modification of 
equipment, nor does it alter the design of plant systems and the 
revised analysis is consistent with the fuel handling accident 
analysis, the proposed change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The margin of safety as defined by 10 CFR Part 100 has not been 
reduced. The calculated dose is a well within of the limits given in 10 
CFR Part 100 or NUREG-0800. The proposed changes do not alter the bases 
for assurance that safety-related activities are performed correctly or 
the basis for any Technical Specification that is related to the 
establishment of or maintenance of a safety margin. Therefore, 
operation of the facility in accordance with the proposed amendments 
would not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: November 27, 2000.
    Description of amendment request: The proposed amendments delete 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated November 27, 2000.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result of 
the TMI-2 accident. The specific intent of the PASS was to provide a 
system that has the capability to obtain and analyze samples of plant 
fluids containing potentially high levels of radioactivity, without 
exceeding plant personnel radiation exposure limits. Analytical results 
of these samples would be used largely for verification purposes in 
aiding the plant staff in assessing the extent of core damage and 
subsequent offsite radiological dose projections. The system was not 
intended to and does not serve a function for preventing accidents and 
its elimination would not affect the probability of accidents 
previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual benefit 
to post accident mitigation. Past experience has indicated that there 
exists in-plant instrumentation and methodologies available in lieu of 
a PASS for collecting and assimilating information needed to assess 
core damage following an accident. Furthermore, the implementation of 
Severe Accident Management Guidance (SAMG) emphasizes accident 
management strategies based on in-plant instruments. These strategies 
provide guidance to the plant staff for mitigation and recovery from a 
severe accident. Based on current severe accident management strategies 
and guidelines, it is determined that the PASS provides little benefit 
to the plant staff in coping with an accident.
    The regulatory requirements for the PASS can be eliminated without 
degrading the plant emergency response. The emergency response, in this 
sense, refers to the methodologies used in ascertaining the condition 
of the reactor core, mitigating the consequences of an accident, 
assessing and projecting offsite releases of radioactivity, and 
establishing protective action recommendations to be communicated to 
offsite authorities. The elimination of the PASS will not prevent an 
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency 
plan (EP), the emergency operating procedures (EOP), and site survey 
monitoring that support modification of emergency plan protective 
action recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any accident 
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated
    The elimination of PASS related requirements will not result in any 
failure mode not previously analyzed. The PASS was intended to allow 
for verification of the extent of reactor core damage and also to 
provide an input to offsite dose projection calculations. The PASS is 
not considered an accident precursor, nor does its existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within the 
containment building.
    Therefore, this change does not create the possibility of a new or 
different kind

[[Page 81922]]

of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The elimination of the PASS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a neutral 
impact to the margin of safety. Methodologies that are not reliant on 
PASS are designed to provide rapid assessment of current reactor core 
conditions and the direction of degradation while effectively 
responding to the event in order to mitigate the consequences of the 
accident. The use of a PASS is redundant and does not provide quick 
recognition of core events or rapid response to events in progress. The 
intent of the requirements established as a result of the TMI-2 
accident can be adequately met without reliance on a PASS.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: November 28, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Table 6.2.1, Minimum Shift Crew 
Composition with Two Separate Control Rooms and TS Section 6.3.1 (2), 
Unit Staff Qualifications for the Shift Technical Advisor (STA). The 
proposed amendments would permit, as an alternative to the current 
dedicated STA, an on-shift senior reactor operator (SRO) position to be 
combined with the required STA position. The proposed amendments would 
require an individual filling either the dedicated STA position or the 
combined SRO/STA position to meet the Technical Specifications 
educational requirements as described in Federal Register Notice 50 FR 
43621, ``Commission Policy Statement on Engineering Expertise on 
Shift.'' These proposed changes are in accordance with the 
recommendations in the NRC Policy Statement on Engineering Expertise on 
Shift, published on October 28, 1985 and transmitted to all power 
reactor licensees and applicants by NRC Generic Letter 86-04, of the 
same title as the October 28, 1985 policy statement, dated February 13, 
1986. As permitted by the policy statement, FPL proposes to exercise 
either of the STA options on a shift-by-shift basis.
    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.
    Implementation of the proposed changes will not involve any 
physical changes to plant systems, structures, or components (SSC), or 
the manner in which these SSCs are operated, maintained, modified, 
tested, or inspected. Therefore, the proposed use of either the dual 
role SRO/STA position or the current dedicated STA position does not 
increase the probability of an accident previously evaluated. 
Implementation of the proposed changes will result in personnel with 
enhanced operational knowledge being assigned to perform the STA 
function of providing accident assessment expertise, and analyzing and 
responding to off normal occurrences when needed.
    The NRC stated preference in the October 28, 1985, Policy Statement 
on Engineering Expertise on Shift, indicates that the NRC has concluded 
that the individual filling the dual role SRO/STA position may perform 
these functions better than a non-licensed individual filling the STA 
position, even when the SRO/STA is concurrently functioning as one of 
the required shift SROs. Therefore, the proposed TS changes do not 
increase the consequences of an accident previously evaluated.
    (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 amendments will not change the physical plant or the 
modes of plant operation defined in the facility license for either St. 
Lucie unit. Changes proposed for the administrative controls do not 
involve the addition or modification of equipment, nor do they alter 
the design or operation of plant systems. Therefore, operation of 
either facility in accordance with its proposed amendments would not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed amendments revise certain administrative controls 
involving the on-site programmatic process for review and approval of 
plant procedures. Neither the scope, nor the requirement to establish, 
maintain, and implement procedures for activities that could affect 
nuclear safety are being changed.
    The NRC stated preference in the October 28, 1985, Policy Statement 
on Engineering Expertise on Shift, indicates that the NRC has concluded 
that the individual filling the dual role SRO/STA position may perform 
these functions better than a non-licensed individual filling the STA 
position, even when the SRO/STA is concurrently functioning as one of 
the required shift SROs. Therefore, the proposed amendments should 
involve an enhancement in a margin on safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: October 30, 2000.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 5.3.2 for Turkey Point Units 3 and 4 to 
extend the residual heat removal (RHR) pump allowed outage time (AOT) 
from 72 hours to 7 days to restore an inoperable

[[Page 81923]]

RHR pump to operable status. The proposed extension is based on the 
projected time required to replace a leaking or failed pump shaft seal, 
perform post-maintenance testing, and complete any additional 
corrective actions that may be needed to restore the pump to operable 
status. The extended RHR pump AOT will provide adequate time so that 
future seal repair activities are completed successfully in a safe 
manner.
    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 RHR system is part of the Emergency Core Cooling System. 
Inoperable RHR pumps are not accident initiators in any accident 
previously evaluated, and an extended AOT to restore operability of an 
inoperable RHR pump would not increase the probability of occurrence of 
accidents previously analyzed. Therefore, this change does not involve 
an increase in the probability of an accident previously evaluated.
    The RHR system is primarily designed to mitigate the consequences 
of the large Loss Of Coolant Accident (LOCA). In addition, the RHR 
system provides for primary system heat removal during unit shutdown 
conditions. The proposed changes do not affect any of the assumptions 
relative to accident initiators or accident response provided in the 
plant safety analyses. Accordingly, the consequences of accidents 
previously evaluated do not change.
    A Probabilistic Safety Assessment (PSA) was performed to evaluate 
the impact of extending the allowed outage time on the RHR pump from 72 
hours to 7 days. FPL concluded from the results of that assessment that 
the risk contribution of the AOT extension is very small, and that the 
net impact of the proposed amendment may be risk neutral.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different kind 
of accident from any previously evaluated.
    The proposed change does not alter the design, physical 
configuration, or modes of operation of the plant. Plant configurations 
that are prohibited by Technical Specifications will not be created by 
the AOT extension. Therefore, the proposed activity does not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The margin of safety associated with the Emergency Core Cooling 
System is established by acceptance criteria for system performance 
defined in 10 CFR 50.46. The proposed amendments will not change these 
acceptance criteria or the operability requirements for equipment that 
is used to achieve such performance as demonstrated in the plant safety 
analyses. Moreover, a Probabilistic Safety Assessment of the risk 
impact of extending the AOT for a single inoperable RHR pump has 
concluded that the risk contribution is very small, RHR system 
reliability can potentially be improved, and the net impact of the 
proposed change may be risk neutral. Therefore, the change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: December 6, 2000.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated December 6, 2000.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result of 
the TMI-2 accident. The specific intent of the PASS was to provide a 
system that has the capability to obtain and analyze samples of plant 
fluids containing potentially high levels of radioactivity, without 
exceeding plant personnel radiation exposure limits. Analytical results 
of these samples would be used largely for verification purposes in 
aiding the plant staff in assessing the extent of core damage and 
subsequent offsite radiological dose projections. The system was not 
intended to and does not serve a function for preventing accidents and 
its elimination would not

[[Page 81924]]

affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual benefit 
to post accident mitigation. Past experience has indicated that there 
exists in-plant instrumentation and methodologies available in lieu of 
a PASS for collecting and assimilating information needed to assess 
core damage following an accident. Furthermore, the implementation of 
Severe Accident Management Guidance (SAMG) emphasizes accident 
management strategies based on in-plant instruments. These strategies 
provide guidance to the plant staff for mitigation and recovery from a 
severe accident. Based on current severe accident management strategies 
and guidelines, it is determined that the PASS provides little benefit 
to the plant staff in coping with an accident.
    The regulatory requirements for the PASS can be eliminated without 
degrading the plant emergency response. The emergency response, in this 
sense, refers to the methodologies used in ascertaining the condition 
of the reactor core, mitigating the consequences of an accident, 
assessing and projecting offsite releases of radioactivity, and 
establishing protective action recommendations to be communicated to 
offsite authorities. The elimination of the PASS will not prevent an 
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency 
plan (EP), the emergency operating procedures (EOP), and site survey 
monitoring that support modification of emergency plan protective 
action recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any accident 
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated.
    The elimination of PASS related requirements will not result in any 
failure mode not previously analyzed. The PASS was intended to allow 
for verification of the extent of reactor core damage and also to 
provide an input to offsite dose projection calculations. The PASS is 
not considered an accident precursor, nor does its existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within the 
containment building.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The elimination of the PASS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a neutral 
impact to the margin of safety. Methodologies that are not reliant on 
PASS are designed to provide rapid assessment of current reactor core 
conditions and the direction of degradation while effectively 
responding to the event in order to mitigate the consequences of the 
accident. The use of a PASS is redundant and does not provide quick 
recognition of core events or rapid response to events in progress. The 
intent of the requirements established as a result of the TMI-2 
accident can be adequately met without reliance on a PASS.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment requests: November 15, 2000.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specification (TS) 3.2.6, ``Allowable Power 
Level--APL,'' and TS 1.38, ``Allowable Power Level (APL),'' definitions 
of APL to remove a condition that limits APL to 100 percent of rated 
thermal power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    No new accident initiators or precursors are created by the 
proposed T/S changes. Reactor thermal power and power distribution 
within the reactor core are not initiators or precursors to any 
previously evaluated accident. There are no physical changes to the 
plant associated with the proposed T/S changes that would create any 
new accident initiators or precursors. Therefore, the proposed T/S 
changes do not increase the probability of occurrence of any accident 
previously evaluated.
    Reactor thermal power up to the calculated value of APL ensures 
that the accident analysis results are not impacted by maintaining 
reactor core power distribution within prescribed limits. Since T/S 1.3 
still contains a limitation on the maximum reactor thermal power 
allowed during normal operations, the normal overall operating limits 
for the reactor core are not changed. Accident analyses generally 
include a calorimetric error allowance of 2% or assume an initial power 
level of at least 102%. Using the additional limit on reactor thermal 
power based on APL ensures operation within the power distribution 
limits assumed in the accident analyses. Therefore, the proposed T/S 
changes do not affect operation of the reactor core and do not modify 
either the maximum acceptable reactor thermal power or the maximum 
allowed power distribution limits.
    The proposed T/S changes do not change or alter the design criteria 
for the systems or components used to mitigate the consequences of any 
design basis accident. The reactor protection system (RPS), including 
reactor trips based upon overall reactor thermal power and power 
distribution within the reactor core, are not affected by the proposed 
T/S changes. The initial conditions of the accident analyses, including 
maximum reactor thermal power and worst-case power distribution within 
the reactor core, are not changed. As a result, the expected operation 
of the emergency core cooling systems (ECCS) are not affected by the 
proposed T/S changes. Radiological consequences of previously evaluated 
accidents are not increased, since overall reactor thermal power and 
power distribution limits are still maintained within the assumptions 
of the accident analyses, and operation of

[[Page 81925]]

the RPS and ECCS is not affected. Therefore, the proposed changes do 
not increase the consequences of any accident and do not impact offsite 
dose considerations.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Reactor thermal power and power distribution within the reactor 
core cannot be an initiator or precursor to an accident. There are no 
physical changes to the plant associated with the proposed T/S changes 
that would create any new accident initiators or precursors. The 
proposed T/S changes do not degrade the reliability of any existing 
system, structure, or component. No new failure modes, malfunctions, or 
system interactions are created. The maximum steady state reactor core 
power level as defined by T/S 1.3 is not changed. The actual power 
distribution limits are not changed since the calculated value of APL 
is not changed. Therefore, the accident analyses assumptions and 
results are unchanged.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    The proposed T/S changes do not change either the overall maximum 
reactor thermal power allowed, or the reactor core power distribution 
limits allowed. Maximum reactor thermal power remains limited by T/S 
1.3. The calculated value of APL in T/S 3.2.6 is not changed, and 
remains as a control to ensure reactor core power distribution limits 
consistent with the accident analyses are satisfied. Therefore, safety 
margins related to power distribution limits are not affected. The 
proposed T/S changes do not affect any of the T/S safety limits or T/S 
limiting safety system settings, and RPS setpoints as defined by the T/
S are not changed or affected.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: November 28, 2000.
    Description of amendment request: The proposed amendment would 
establish technical specifications (TSs) for the emergency service 
water system. It would also revise TS 3.0 to include general 
requirements for system operability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The EFT-ESW [emergency filtration train-emergency service water] 
System is not an accident initiator. The proposed amendment provides 
operability requirements and surveillance requirements to ensure the 
ESW System is available and operable when required for accident 
mitigation. The proposed operability requirements and allowed outage 
times are consistent with similar requirements for the systems 
supported by the EFT-ESW System. Dose to the public and the Control 
Room operators are not affected by the proposed change. The proposed 
general LCO [limiting condition for operation] provides direction with 
respect to actions to be taken when support systems are inoperable.
    The proposed Technical Specification change does not introduce new 
equipment operating modes, nor does the proposed change alter existing 
system relationships. The proposed amendment does not introduce new 
failure modes.
    Therefore, the proposed amendment will not significantly increase 
the probability or the consequences of an accident previously 
evaluated.
    The proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    The proposed Technical Specification change does not introduce new 
equipment operating modes, nor does the proposed change alter existing 
system relationships. The proposed amendment does not introduce new 
failure modes. The proposed amendment does not alter the equipment 
required for accident mitigation and considers the effects on supported 
systems when a support system is inoperable. When support systems are 
inoperable, actions are specified to be taken consistent with safe 
plant operation.
    Therefore, the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment will not involve a significant reduction in 
the margin of safety.
    The proposed amendment provides specifications for the EFT-ESW 
System which are consistent with current Technical Specification 
requirements for other equipment. The proposed changes ensure that the 
EFT-ESW and other support systems will be available when required and 
provides adequate alternative actions when the support systems are not 
available. The allowed outage times for the EFT-ESW Pumps are 
consistent with that allowed for other equipment that would have 
similar importance to accident mitigation. The proposed general LCO 
does not result in a significant reduction in the margin of safety 
since it imposes requirements already in technical specifications for 
support systems included in technical specifications. In cases where 
support systems [are] not included in technical specifications, the 
proposed general LCO does not apply and actions determined to be 
required by the technical specifications will be taken for the 
supported systems.
    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 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: September 5, 2000.
    Description of amendment request: The proposed amendment would 
revise the Fort Calhoun Station, Unit No. 1 (FCS) Technical 
Specifications (TS) to change the definition section, TS

[[Page 81926]]

Sections 2.10, 3.10, and 5.9, and the Bases of TS 1.1 and 1.3, to allow 
the use of nuclear fuel fabricated by Siemens Power Corporation at FCS. 
The definition of unrodded planar radial peaking factor 
(Fxy) and TS 2.10.4(3) are being deleted and TS 3.10 is 
being revised to reflect the deletion of this peaking factor. TS 5.9.5 
is being revised to incorporate NRC-approved methodologies necessary to 
determine core operating limits with nuclear fuel from Siemens Power 
Corporation. The Bases to TS 1.1 and 1.3 are being revised to delete 
the discussion of the CE-1 correlation that is currently used to 
calculate minimum departure from nucleate boiling ratio and the value 
calculated by this method.
    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 amendment is to incorporate Siemens Power Corporation 
topical reports for conducting reload analyses that have been 
previously reviewed and approved by the NRC. The applicable FCS 
Technical Specifications (TS) supported by these topical reports are 
being revised. These changes are necessary to support using nuclear 
fuel supplied by Siemens Power Corporation.
    It is proposed to revise the Bases of TS 1.1 and 1.3 to reflect 
changes in methodologies for calculating the minimum Departure from 
Nucleate Boiling Ration (DNBR). The proposed methodology for 
determining the minimum DNBR for fuel supplied by Siemens Power 
Corporation is the NRC-approved EMF-92-153(P)(A) and Supplement 1, HTP: 
Departure from Nucleate Boiling Correlation for High Thermal 
Performance Fuel. As stated in the Basis of TS 1.1, Fort Calhoun 
Station currently uses the NRC-approved CE-1 correlation with a minimum 
DNBR value of 1.18, which provides a 95% probability at a 95% 
confidence level that DNB will not occur for any operating condition. 
For Siemens fuel, using the HTP correlation with a minimum DNBR of 
1.14, as proposed, will continue to provide a 95% probability at a 95% 
confidence level that DNB will not occur during any operating 
condition. The CE-1 correlation is more restrictive than the HTP 
correlation that will be used to predict the minimum DNBR limits for 
the Siemens fuel. For a given set of reactor coolant conditions, the 
CE-1 correlation provides a lower critical heat flux than the HTP 
correlation. Therefore, this change will not significantly increase the 
probability or consequences of an accident previously evaluated.
    It is proposed that the total planar radial peaking factor, 
FxyT, be eliminated from the Technical 
Specifications. The current need for this parameter is to protect 
assumptions about the maximum amount of planar peaking in the core. The 
limitation on the total planar radial peaking factor, 
FxyT, is provided to ensure that the assumptions 
used in the analysis for establishing the Linear Heat Rate and Local 
Power Density--High, Limiting Conditions for Operation, and Limiting 
Safety Systems Settings set-points remain valid during operation. In a 
two-dimensional set-point analysis, as currently conducted, 
FxyT is combined with the maximum axial power 
profile (Fz) to produce the maximum allowable peaking factor 
(Fq) or equivalent Linear Heat Rate. This ensures 
conservative operation relative to assumptions on linear heat rate used 
as input to the loss of coolant accident and other transient analyses. 
In a three-dimensional analysis, as proposed with the use of Siemens 
methodology, these peaks are calculated directly during a series of 
pre-determined maneuvers (axial shape oscillation, power maneuver, or 
other transient).
    Direct calculation of these peaks negates the need to make 
inferences about the amount of planar radial peaking that occurs in any 
particular plane within the core. Therefore, this change will not 
significantly increase the probability or consequences of an accident 
previously evaluated.
    It is proposed to add NRC-approved methodologies from Siemens Power 
Corporation to TS that are necessary to evaluate core parameters. The 
proposed additions of NRC-approved topical reports to the TS do not 
modify the manner in which the topical reports may be implemented. The 
core operating limits will continue to be determined using NRC-approved 
analytical methods. The plant will continue to operate within the 
limits specified by the Core Operating Limits Report and will take 
corrective actions as required by the current Technical Specifications 
should these limits be exceeded. Therefore, these changes will not 
significantly increase 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.
    No new or different modes of operation are proposed as a result of 
these changes. The proposed revisions do not change any equipment 
required to mitigate the consequences of an accident. The proposed 
additions of NRC-approved topical reports to the TS do not modify the 
manner in which the topical reports may be implemented. The plant will 
continue to operate within the limits specified by the Core Operating 
Limits Report and will take corrective actions as required should these 
limits be exceeded. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    As required by TS 5.9.5, the analytical methods used to determine 
the core operating limits shall be those previously reviewed and 
approved by the NRC. The proposed changes incorporate methodologies 
applicable for use with fuel supplied by Siemens Power Corporation that 
have been approved by the NRC as documented by Safety Evaluation 
Reports. Technical Specification 5.9.5 also requires that the core 
operating limits shall be determined so that all applicable limits of 
the safety analysis are met. These requirements will continue to be 
met. Therefore, OPPD concludes that 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: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 18, 2000.
    Description of amendment request: The proposed amendment would 
revise the Fort Calhoun Station Unit 1 (FCS) Technical Specifications 
(TSs) to (1) extend the validity of the existing TS Figure 2-1A (RCS 
[reactor coolant system] Pressure-Temperature Limits for Heatup) and 
Figure 2-1B (RCS Pressure-Temperature Limits for Cooldown) from

[[Page 81927]]

20.0 effective full power years (EFPY) to 24.25 EFPY, (2) delete Figure 
2-3 (Predicted Radiation Induced NDTT [nil ductility transition 
temperature] Shift), and (3) provide replacement guidance in TSs 
2.1.2(6)(a) and (b) for use of the most current fluence analysis and 
Regulatory Guide 1.99, Revision 2, ``Radiation Embrittlement of Reactor 
Vessel Materials,'' for projecting reference temperature nil ductility 
(RTNDT) at 24.25 EFPY. The proposed amendment would also 
revise the associated Bases section of TS 2.1.2.
    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 NRC previously approved Technical Specification Amendment No. 
161 in March 1994 for the use of RCS Pressure-Temperature (P-T) Limits 
good to 20.0 EFPY. The proposed changes in this submittal reflect the 
validity of these same curves from 20.0 EFPY to 24.25 EFPY based on the 
implementation of extreme low radial leakage fuel management in 1992 
(Cycle 14). Significant reductions in the fast neutron flux to the 
limiting 3-410 axial weld in the Fort Calhoun Station reactor pressure 
vessel were obtained, thus significantly increasing the time to when 
the fast neutron fluence input to the derivation of the previously 
approved P-T curves will be reached. Since no inputs (including assumed 
material properties of the limiting weld) to the existing analysis are 
being changed, extension of the validity of the curves from 20.0 EFPY 
to 24.25 EFPY is justified. In addition, deletion of Figure 2-3 and 
references to it are proposed. This proposed change removes an outdated 
figure which is non-operational in nature. The application of the 
current Regulatory Guide 1.99, Revision 2 is more appropriate for these 
purposes. Administrative changes to the Basis section of TS 2.1.2 are 
proposed to reflect the extension to 24.25 EFPY.
    No accidents previously analyzed are affected by these changes, and 
it can be concluded that there is no significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not physically alter the configuration of 
the plant and no new or different mode of operation is proposed. 
Extending the validity of the P-T curves more accurately projects 
reactor vessel embrittlement by accounting for improvements in FCS fuel 
management which have significantly reduced the fast neutron fluence to 
the limiting 3-410 axial weld, incorporates improved operating cycle 
efficiency, and applies the WCAP-15443, Revision 0 fluence analysis. 
The revised fluence analysis uses the ENDF/B-VI Nuclear Cross Section 
Library. Deletion of Figure 2-3 represents a change which does not 
affect plant operations. Figure 2-3 is administrative in nature, and 
proposed revisions to Specifications 2.1.2(6)(a) and (b) provide 
guidance consistent with the current Regulatory Guide for P-T curves 
updates. Update of the Technical Specification 2.1.2 Basis section 
represents an administrative change that does not affect plant 
operation.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes to extend the validity of Technical 
Specification Figures 2-1A and 2-1B to 24.25 EFPY are consistent with 
the extreme low radial leakage fuel management implemented in 1992 
(Cycle 14) and performance/application of the updated fluence analysis 
described above. With no changes to the inputs of the existing P-T 
limits analysis, there is no reduction in the margin of safety. Figure 
2-3 is not used to provide limits on plant operation, and deletion of 
this figure, which uses a pre-Regulatory Guide 1.99, Revision 2 
embrittlement correlation, is considered an improvement in the 
consistency of the requirements outlined in the Technical 
Specifications. This Figure is not used in plant operation and provides 
only a general indication of the RTNDT shift. The TS 2.1.2 
Basis section changes are administrative in nature and do not affect 
the margin of safety. The changes serve to maintain consistency with 
the NRC approval of Amendment No. 161.
    In summary, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 27, 2000.
    Description of amendment request: The proposed amendment would 
revise Section 3.7 of the Fort Calhoun Station Unit 1 Technical 
Specifications to eliminate item 3.7(4) ``13.8 Kv Transmission Line'' 
which states: ``The 13.8 Kv transmission line will be energized and 
loaded to minimum shutdown requirements at each refueling outage 
following installation.''
    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.
    Eliminating the 13.8 kV testing requirement would have no impact 
upon the probability of an accident previously evaluated. The circuit 
breaker connecting the 13.8 kV power supply to the station electrical 
busses is normally open, so this power supply could not play a role in 
the initiation of any accident.
    Eliminating the 13.8 kV testing requirement would have no impact 
upon the consequences of an accident previously evaluated. Existing 
accident analyses take no credit for the 13.8 kV power supply.
    The 13.8 kV power supply is not credited for mitigation of 
licensing basis transients or postulated events added to the USAR 
[Updated Safety Analysis Report] by NRC requirements, such as Station 
Blackout (SB0).
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The 13.8 kV power supply is only capable of supplying a limited 
number of components in the unlikely event that 161 kV, 345 kV, and the 
diesel-generators are unavailable. Eliminating the 13.8 kV testing 
requirement would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.

[[Page 81928]]

    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    Testing of the 13.8 kV power supply, as described in Technical 
Specification 3.7(4), is unrelated to any margin of safety. Therefore, 
deletion of the testing requirement will not reduce 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: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment requests: November 30, 2000.
    Description of amendment requests: The proposed license amendments 
would change Technical Specification Section 3.5.1, ``Accumulators,'' 
by revising the limits for accumulator borated water volume 
(Surveillance Requirement (SR) 3.5.1.2) and nitrogen cover pressure (SR 
3.5.1.3) to reflect analysis limits. These TS currently reflect nominal 
limits. These amendments are revising TS values consistent with other 
similar TS parameters which will aid in future clarity.
    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 accumulators only function following an accident. They cannot 
initiate an accident. The proposed changes have no impact to plant 
operation and are administrative in nature. Changing the technical 
specification (TS) limits for accumulator volume and pressure from 
nominal to analysis values will provide greater consistency within the 
TS. Changing the volume limits to cubic feet verse[u]s percent level 
will eliminate any potential for future revision of these limits 
because of instrument tap relocation.
    Plant parameters will continue to be administratively controlled 
within the allowed analysis parameters. The proposed limits for tank 
volume and nitrogen cover pressure are consistent with analysis values 
documented in the Final Safety Analysis Report and assume that the 
accident consequences remain unchanged.
    There are no hardware changes or changes in the method by which any 
safety-related plant system performs its safety function.
    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 accumulators only function following an accident. They cannot 
initiate an accident. The proposed changes are administrative in 
nature.
    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 changes are administrative in nature so there are no new 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures are [sic] introduced.
    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.
    The proposed changes are administrative in nature.
    The proposed changes do not affect the acceptance criteria for any 
analyzed event. There will be no effect on the manner in which safety 
limits or limiting safety system settings are determined nor will there 
be any effect on those plant systems necessary to assure the 
accomplishment of protection functions.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment requests: November 30, 2000.
    Description of amendment requests: The proposed license amendments 
would change the administrative controls sections of Technical 
Specification (TS) 5.5.14b and 5.5.14b.2 to incorporate the changes 
made to 10 CFR Part 50, Section 50.59. The proposed amendments would 
replace the word ``involve'' with ``require'' in TS 5.5.14b and revise 
TS 5.5.14b.2 to delete the reference to ``unreviewed safety question'' 
and restate the requirement as ``a change to the updated Final Safety 
Analysis Report or Bases that requires NRC approval pursuant to 10 CFR 
50.59.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change replaces the word ``involve'' with ``require'' 
and deletes reference to the term ``unreviewed safety question'' 
consistent with 10 CFR [Part 50, Section] 50.59. Deletion of the term 
``unreviewed safety question'' was approved by the NRC with the 
revision to 10 CFR 50.59. Consequently, the probability of an accident 
previously evaluated is not significantly increased. Changes to the 
Technical Specification (TS) Bases are still evaluated in accordance 
with 10 CFR 50.59. As a result, the consequences of any accident 
previously evaluated are not significantly affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing plant operation. These changes are 
considered administrative changes and do not modify, add, delete,

[[Page 81929]]

or relocate any technical requirements in the TS.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes will not reduce the margin of safety because 
they have no effect on any safety analyses assumptions. Changes to the 
TS Bases that result in meeting the criteria in paragraph (c)(2) of 10 
CFR 50.59 will still require NRC approval. The proposed changes to TS 
5.5.14 are considered administrative in nature based on the revision to 
10 CFR 50.59.
    Therefore, the proposed changes do not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 25, 2000.
    Description of amendment request: The amendment revises the Action 
Statements associated with Technical Specifications (TSs) Table 
3.3.7.5-1 (``Accident Monitoring Instrumentation'') concerning the 
Drywell Hydrogen/Oxygen (H2/O2) Concentration 
Analyzers, and the associated TS Bases. PECO Energy proposes to add new 
Action Statements 82a and 82b concerning channel operability, which 
will replace the current requirements of Action Statements 80a and 80b, 
respectively, for the Drywell Hydrogen/Oxygen Concentration Analyzers.
    Under the existing TS Action Statements for Table 3.3.7.5-1 
(``Accident Monitoring Instrumentation''), with the number of operable 
accident monitoring instrumentation channels less than the ``required'' 
number of channels (quantity 2), restore the inoperable channels within 
7 days or be in at least hot shutdown within the following 12 hours 
(Action Statement 80a). Additionally, with the number of operable 
accident monitoring instrumentation channels less than the ``minimum'' 
number of channels (quantity 1), restore the inoperable channel(s) 
within 48 hours or be in at least hot shutdown within the following 12 
hours (Action Statement 80b).
    Proposed Action Statement 82a for Table 3.3.7.5-1 will extend the 
duration from 7 to 30 days for less than the ``required'' number 
operable of channels. Additionally, the proposed Action Statement 82a 
will require that if the operable channel(s) cannot be restored within 
the 30 days, then a Special Report shall be provided to the NRC within 
the following 14 days.
    Proposed Action 82b for Table 3.3.7.5-1 will extend the duration 
from 48 hours to 72 hours for less than the ``minimum'' number of 
operable channels. If the inoperable channel(s) cannot be restored with 
the 72 hours, then be in hot shutdown with the next 12 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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c).
    1. The proposed [technical specification] TS changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed TS changes modify the Action Statements associated 
with the duration that the Drywell Hydrogen/Oxygen Concentration 
Analyzers can be inoperable. The Drywell Hydrogen/Oxygen Concentration 
Analyzers are not accident initiating equipment and are monitoring 
devices required to be available for monitoring hydrogen and oxygen 
following a LOCA. These analyzers do not perform any automatic or 
control functions. Therefore, the proposed changes will not increase 
the probability of an accident previously evaluated.
    In the event of a failure of the Drywell Hydrogen/Oxygen 
Concentration Analyzers following a LOCA, concentrations of hydrogen 
and oxygen can be measured by utilizing grab samples with the post-
accident sampling system. A single failure of either analyzer package 
would render that affected package inoperable with the redundant 
package fully capable of performing the required function at full 
capacity. Following a postulated LOCA, the hydrogen recombiners will be 
utilized to ensure that the oxygen concentration in the primary 
containment is maintained below the lower flammability limit as 
required by plan emergency procedures.
    The extended completion times are based on the passive nature of 
the instrument (no critical automatic action is assumed to occur from 
these instruments), the low probability of an event requiring post-
accident instrumentation during this interval, and the availability of 
alternate means to obtain the required information. Therefore, the 
proposed TS changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed technical specification changes modify the Action 
Statements associated with the duration that the Drywell Hydrogen/
Oxygen Concentration Analyzers can be inoperable. They do not change 
the design or configuration of the plant. The Drywell Hydrogen/Oxygen 
Concentration Analyzers are not accident initiating equipment, and are 
monitoring devices required to be available for monitoring hydrogen and 
oxygen following a LOCA. The proposed changes do not create a system-
level failure mode different than those that already exist. In 
addition, there are no operation or failure modes of the Drywell 
Hydrogen/Oxygen Concentration Analyzers that are accident initiators. 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes in Action Statements do not affect any safety 
limits or analytical limits. There are also no changes to accident of 
transient core thermal hydraulic conditions, minimum combustible 
concentration limits, or fuel or reactor coolant boundary design 
limits, as a result of these proposed changes. The proposed Technical 
Specification changes modify the Action Statements associated with the 
duration that the Drywell Hydrogen/Oxygen Concentration Analyzers can 
be inoperable. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.
    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.

[[Page 81930]]

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

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

    Date of amendment request: November 29, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to reflect the enabling of the 
Oscillation Power Range Monitor (OPRM) instrumentation reactor 
protection system (RPS) trip function. The OPRM is designed to detect 
the onset of reactor core power oscillations resulting from thermal-
hydraulic instability and suppresses them by initiating a reactor scram 
via the RPS trip logic.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change specifies limiting conditions for operations, 
required actions and surveillance requirements of the OPRM system and 
allows operation in regions of the power to flow map currently 
restricted by the requirements of Interim Corrective Actions (ICAs) and 
certain limiting conditions of operation of Technical Specifications 
(TS) 3.4.1. The OPRM system can automatically detect and suppress 
conditions necessary for thermal-hydraulic (T-H) instability. A T-H 
instability event has the potential to challenge the Minimum Critical 
Power Ratio (MCPR) safety limit. The restrictions of the ICAs and TS 
3.4.1 were imposed to ensure adequate capability to detect and suppress 
conditions consistent with the onset of T-H oscillations that may 
develop into a T-H instability event. With the installation of the OPRM 
System, these restrictions are no longer required.
    The probability of a T-H instability event is most significantly 
impacted by power to flow conditions such that only during operation 
inside specific regions of the power to flow map, in combination with 
power shape and inlet enthalpy conditions, can the occurrence of an 
instability event be postulated to occur. Operation in these regions 
may increase the probability that operation with conditions necessary 
for a T-H instability can occur.
    However, when the OPRM is operable with operating limits as 
specified in the COLR [Core Operating Limits Report], the OPRM can 
automatically detect the imminent onset of local power oscillations and 
generate a trip signal. Actuation of an RPS trip will suppress 
conditions necessary for T-H instability and decrease the probability 
of a T-H instability event. In the event the trip capability of the 
OPRM is not maintained, the proposed change includes actions which 
limit the period of time before the effected OPRM channel (or RPS 
system) must be placed in the trip condition. If these actions would 
result in a trip function, an alternate method to detect and suppress 
thermal hydraulic oscillations is required. In either case the duration 
of this period of time is limited such that the increase in the 
probability of a T-H instability event is not significant. Therefore 
the proposed change does not result in a significant increase in the 
probability of an accident previously evaluated.
    An unmitigated T-H instability event is postulated to cause a 
violation of the MCPR safety limit. The proposed change ensures 
mitigation of T-H instability events prior to challenging the MCPR 
safety limit if initiated from anticipated conditions by detection of 
the onset of oscillations and actuation of an RPS trip signal. The OPRM 
also provides the capability of an RPS trip being generated for T-H 
instability events initiated from unanticipated but postulated 
conditions. These mitigating capabilities of the OPRM system would 
become available as a result of the proposed change and have the 
potential to reduce the consequences of anticipated and postulated T-H 
instability events. Therefore, the proposed change does not 
significantly increase the consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change specifies limiting conditions for operations, 
required actions and surveillance requirements of the OPRM system and 
allows operation in regions of the power to flow map currently 
restricted by the requirements of ICAs and TS 3.4.1. The OPRM system 
uses input signals shared with APRM and rod block functions to monitor 
core conditions and generate an RPS trip when required. Quality 
requirements for software design, testing, implementation and module 
self-testing of the OPRM system provide assurance that no new equipment 
malfunctions due to software errors are created. The design of the OPRM 
system also ensures that neither operation nor malfunction of the OPRM 
system will adversely impact the operation of other systems and no 
accident or equipment malfunction of these other systems could cause 
the OPRM system to malfunction or cause a different kind of accident. 
Therefore, operation with the OPRM system does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Operation in regions currently restricted by the requirements of 
ICAs and TS 3.4.1 is within the nominal operating domain and ranges of 
plant systems and components for which postulated equipment and 
accidents have been evaluated. Therefore operation within these regions 
does not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change which specifies limiting conditions for 
operations, required actions and surveillance requirements of the OPRM 
system and allows operation in certain regions of the power to flow map 
does not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change specifies limiting conditions for operations, 
required actions and surveillance requirements of the OPRM system and 
allows operation in regions of the power to flow map currently 
restricted by the requirements of ICAs and TS 3.4.1.
    The OPRM system monitors small groups of LPRM signals for 
indication of local variations of core power consistent with T-H 
oscillations and generates an RPS trip when conditions consistent with 
the onset of oscillations are detected. An unmitigated T-H instability 
event has the potential to result in a challenge to the MCPR safety 
limit. The OPRM system provides the capability to automatically detect 
and suppress conditions which might result in a T-H instability event 
and thereby maintains the margin of safety by providing automatic 
protection for the MCPR safety limit while significantly reducing the 
burden on the control room operators. In the event the trip capability 
of the OPRM is not maintained, the proposed change includes actions 
which limit the period of time before the effected OPRM channel (or RPS 
system) must be placed in the trip condition. If these actions

[[Page 81931]]

would result in a trip function, an alternate method to detect and 
suppress thermal hydraulic oscillations is required. Since, in either 
case, the duration of this period of time is limited so that the 
increase in the probability of a T-H instability event is not 
significant. Operation with the OPRM system does not involve a 
significant reduction in a margin of safety.
    Operation in regions currently restricted by the requirements of 
ICAs and TS 3.4.1 is within the nominal operating domain assumed for 
identifying the range of initial conditions considered in the analysis 
of anticipated operational occurrences and postulated accidents. 
Therefore, operation in these regions does not involve a significant 
reduction in the margin of safety.
    The proposed change, which specifies limiting conditions for 
operations, required actions and surveillance requirements of the OPRM 
system and allows operation in certain regions of the power to flow 
map, does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of application request: November 21, 2000 (ULNRC-04346)
    Description of amendment request: The proposed amendment request 
would change Table 3.3.2-1, ``Engineered Safety Feature Actuation 
System Instrumentation,'' of the Technical Specifications. The change 
would add Surveillance Requirement (SR) 3.3.2.10 to the SRs for the 
following two engineered safety feature actuation system (ESFAS) 
instrumentation in the table: item f, loss of offsite power, and item 
h, auxiliary feedwater pump suction transfer on suction pressure--low. 
The licensee also identified that there would be changes to the Final 
Safety Analysis Report (FSAR).
    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.
    Overall protection system performance will remain within the bounds 
of the previously performed accident analyses since there are no 
hardware changes. The Reactor Trip System (RTS) and Engineered Safety 
Feature Actuation System (ESFAS) instrumentation will be unaffected. 
These protection systems 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 request are 
maintained.
    The proposed change imposes more stringent surveillance testing 
requirements to ensure safety-related structures, systems, and 
components are tested in a manner consistent with the safety analysis 
and licensing basis.
    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.
    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. This change will not affect the normal method of plant 
operation or change any operating parameters. No performance 
requirements will be affected; however, the proposed change does impose 
additional surveillance testing requirements. These additional 
requirements are consistent with assumptions made in the safety 
analysis and licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result of 
this change. There will be no adverse effect or challenges imposed on 
any safety-related system as a result of this change.
    This change does not alter the design or performance of the 7300 
Process Protection System, Nuclear Instrumentation System, or Solid 
State Protection System used in the plant protection systems.
    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 (DNBR) limits, heat flux hot 
channel factor (FQ), nuclear enthalpy rise hot channel 
factor (FdeltaH), 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 
[NRC] Standard Review Plan [(NUREG-0800)] will continue to be met.
    The imposition of more stringent surveillance requirements [in the 
change] increase the margin of safety by ensuring that the affected 
safety analysis assumptions on equipment response time are verified on 
a periodic frequency.
    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.

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

    Date of application request: November 22, 2000.
    Description of amendment request: The proposed change to Callaway 
Technical Specification (TS) 5.5.14, which ensures that a program 
exists for processing changes to the TS Bases, would replace the word 
``involve'' with ``require'' and deletes the phrase ``unreviewed safety 
question'' as

[[Page 81932]]

defined in 10 CFR Part 50, Section 50.59.
    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 replace the word ``involve'' with ``require'' 
and deletes the phrase ``unreviewed safety question'' as defined in 10 
CFR 50.59. The above changes are consistent with the revision to 10 CFR 
50.59. Consequently, the probability of an accident previously 
evaluated is not significantly increased. Changes to the Technical 
Specification Bases are still evaluated in accordance with 10 CFR 
50.59. As a result, the consequences of any accident previously 
evaluated are not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing plant operation. These changes are 
considered administrative changes and do not modify, add, delete, or 
relocate any technical requirements in the TS.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes will not reduce the margin of safety because 
they have no effect on any safety analyses assumptions. Changes to the 
TS Bases that result in meeting the criteria in paragraph (c)(2) of 10 
CFR 50.59 will still require NRC approval. The proposed changes to TS 
5.5.14 are considered administrative in nature based on the revisions 
to 10 CFR 50.59.
    Therefore, the proposed changes do not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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.

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

    Date of amendment request: December 7, 2000 (ET 00-0041).
    Description of amendment request: The proposed amendment request 
would change Table 3.3.2-1, ``Engineered Safety Feature Actuation 
System Instrumentation,'' of the Technical Specifications (TSs). The 
change would add Surveillance Requirement (SR) 3.3.2.10 to the SRs for 
the following two engineered safety feature actuation system (ESFAS) 
instrumentation in the table: item 6.f, loss of offsite power, and item 
6.h, auxiliary feedwater pump suction transfer on suction pressure--
low. The licensee also identified that there would be changes to the 
Updated Safety Analysis Report (USAR) and changes to the Bases for the 
TSs.
    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.
    Overall protection system performance will remain within the bounds 
of the previously performed accident analyses since there are no 
hardware changes. The Reactor Trip System (RTS) and Engineered Safety 
Feature Actuation System (ESFAS) instrumentation will be unaffected. 
These protection systems 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 request are 
maintained.
    The proposed change imposes more stringent surveillance testing 
requirements to ensure safety related structures, systems, and 
components are tested in a manner consistent with the safety analysis 
and licensing basis.
    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 
USAR.
    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. This change will not affect the normal method of plant 
operation or change any operating parameters. No performance 
requirements will be affected; however, the proposed change does impose 
additional surveillance testing requirements. These additional 
requirements are consistent with assumptions made in the safety 
analysis and licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result of 
this change. There will be no adverse effect or challenges imposed on 
any safety related system as a result of this change.
    This change does not alter the design or performance of the 7300 
Process Protection System, Nuclear Instrumentation System, or Solid 
State Protection System used in the plant protection systems.
    Therefore, the proposed changes do not create 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.
    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 (DNBR) limits, heat flux hot 
channel factor (FQ), nuclear enthalpy rise hot channel 
factor (FdeltaH), 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 
[NRC] Standard Review Plan

[[Page 81933]]

[(NUREG-0800)] will continue to be met.
    The imposition of more stringent surveillance testing requirements 
[in the change] increases the margin of safety by ensuring that the 
affected safety analysis assumptions on equipment response time are 
verified on a periodic frequency.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

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

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

    Date of application for amendment: March 19, 1999, and supplemented 
by letters dated April 17, May 5, June 16, July 26, and November 21, 
2000.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) 1.40, ``Spent Fuel Pool Storage Pattern''; 1.41, 
``3-OUT-OF-4 AND 4-OUT-OF-4''; 3/4.9.1.2, ``Boron Concentration''; 3/
4.9.7, ``Crane Travel-Spent Fuel Storage Areas''; 3/4.9.13, ``Spent 
Fuel Pool--Reactivity''; 3.9.14, ``Spent Fuel Pool--Storage Pattern''; 
5.6.1.1, ``Design Features--Criticality''; and 5.6.3, ``Design 
Features--Capacity.'' In addition, the amendment revises INDEX pages 
xii and xv for new figures and page numbers and replaces Figures 3.9-1 
and 3.9-2 with four new figures and make changes to the TS Bases 
consistent with changes to their respective TS sections.
    Date of issuance: November 28, 2000.
    Amendment No.: 189.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of individual notice in Federal Register: December 4, 2000 (65 
FR 75736).

Notice of Issuance of Amendments to Facility Operating Licenses

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

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: November 22, 1999, as 
supplemented November 24, 1999 and September 12, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification 5.5.11, ``Ventilation Filter Testing Program'' for 
laboratory testing of charcoal in engineered safety feature ventilation 
systems to reference American Society for Testing and Materials D3803-
1989 ``Standard Test Method for Nuclear-Grade Activated Carbon.''
    Date of issuance: December 7, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 238 and 212.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73085)
    The November 24, 1999, and September 12, 2000, submittals provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated December 7, 2000.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: September 15, 2000.
    Brief description of amendments: The amendments implement Technical 
Specification Task Force (TSTF)-134, Revision 1. TSTF-134 revises 
Technical Specification Surveillance Requirements (SR) 3.1.7.2 which 
verifies control element assembly (CEA) trip function from 50 percent 
withdrawn position, by adding a note allowing SR 3.1.7.2 not be 
performed if TS SR 3.1.4.6 (CEA drop time test) has been met. TSTF-134, 
Revision 1, was approved by the Nuclear Regulatory Commission on April 
21, 1998.
    Date of issuance: December 11, 2000.

[[Page 81934]]

    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 239 and 213.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62384).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated December 11, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 5, 1999, as supplemented on 
December 22, 1999, and September 18, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.5, ``Instrumentation Systems,'' for the reactor 
protection system and engineered safety features actuation system 
instrumentation. Specifically, the amendment: (1) Revises the allowed 
outage times for the instrumentation, (2) allows on-line testing and 
maintenance of instrumentation, and (3) revises the associated Bases 
section. The amendment also includes several editorial changes to TS 
Tables 3.5-2 and 3.5-3.
    Date of issuance: November 30, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 212.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59221).
    The December 22, 1999, and September 18, 2000, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated November 30, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 22, 2000, as supplemented 
on October 3 and 15, 2000.
    Brief description of amendment: The amendment revises: (1) 
Technical Specification (TS) 3.10.4, ``Rod Insertion Limits,'' to allow 
on-line calibration of the rod position indicator (RPI) channels during 
operating cycle 15, and (2) TS 3.10.6, ``Inoperable Rod Position 
Indicator Channels,'' to allow extended RPI deviation limits during 
cycle 15.
    Date of issuance: December 12, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 213.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56948).
    The October 3 and 15, 2000, letters provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 12, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 21, 2000.
    Brief description of amendment: This amendment deleted references 
to stainless steel as the material for reactor coolant system and 
reactor coolant pressure boundary component fasteners from Table 1.8-1 
and 1.8-2 of the Beaver Valley Power Station, Unit No. 1, Updated Final 
Safety Analysis Report (UFSAR).
    Date of issuance: December 4, 2000.
    Effective date: As of date of issuance.
    Amendment No.: 235.
    Facility Operating License No. DPR-66: Amendment authorized changes 
to the UFSAR.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37426).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 4, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: July 19, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) surveillance requirements of the safety-related 
ventilation system charcoal consistent with the actions requested in 
Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated 
Charcoal,'' dated June 3, 1999. Systems impacted include the control 
room emergency ventilation system, the shield building ventilation 
system, the emergency core cooling system area ventilation system, and 
the fuel pool ventilation system--fuel storage.
    Date of Issuance: December 7, 2000.
    Effective Date: December 7, 2000.
    Amendment No.: 167.
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48749).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 7, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: June 21, 2000.
    Brief description of amendments: These amendments relocate 
Technical Specification (TS) Surveillance Requirement 4.8.1.1.2.e.1, 
regarding the emergency diesel generator (EDG) inspection program, to a 
licensee controlled maintenance program that will be incorporated by 
reference into the next revision of the Updated Final Safety Analysis 
Report for each St. Lucie unit. Upon relocation to the licensee 
controlled maintenance program, the effectiveness of the maintenance on 
the EDGs and support systems will be monitored pursuant to the 
Maintenance Rule 10 CFR 50.65.
    Date of Issuance: December 7, 2000.
    Effective Date: December 7, 2000.
    Amendment Nos.: 168 and 111.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the TS.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48750).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 7, 2000.
    No significant hazards consideration comments received: No.

[[Page 81935]]

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

    Date of application for amendment: November 30, 1999, as 
supplemented June 28, 2000, and November 3, 2000.
    Brief description of amendment: The amendment revises Technical 
Specifications Sections 3.7.2, ``Control Room Envelope Filtration 
(CREF) System,'' and 5.5.7, ``Ventilation Filter Testing Program 
(VFTP)'' for laboratory testing of charcoal filters to reference 
American Society for Testing and Materials standard D3803-1989, 
``Standard Test Method for Nuclear-Grade Activated Carbon.''
    Date of issuance: December 1, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days of issuance.
    Amendment No.: 95.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51358).
    The November 3, 2000, submittal did not change the initial proposed 
no significant hazards consideration determination.
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated December 4, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 11 and May 12, 2000, as 
supplemented by letters dated June 13, June 16, July 14, September 21, 
October 26, and November 3, 2000.
    Brief description of amendment: The amendment grants a conforming 
amendment to the License and the Technical Specifications for the 
approval of the transfer of the license for the Indian Point Nuclear 
Generating Unit No. 3 (IP3) held by the Power Authority of the State of 
New York to Entergy Nuclear IP3, LLC. to possess and use IP3 and to 
Entergy Nuclear Operations, Inc. (ENO) to possess, use and operate IP3.
    Date of issuance: November 21, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 203.
    Facility Operating License No. DPR-64: Amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: June 28, 2000 (65 FR 
39954).
    The supplemental information did not expand the scope of the 
application as originally noticed in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 9, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 11 and May 12, 2000, as 
supplemented by letters dated June 13, June 16, July 14, September 21, 
October 26, and November 3, 2000.
    Brief description of amendment: The amendment grants a conforming 
amendment to the License and the Technical Specifications for the 
approval of the transfer of the license for the James A. FitzPatrick 
Nuclear Power Plant (FitzPatrick) held by the Power Authority of the 
State of New York to Entergy Nuclear FitzPatrick, LLC. to possess and 
use FitzPatrick and to Entergy Nuclear Operations, Inc. (ENO) to 
possess, use and operate FitzPatrick.
    Date of issuance: November 21, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 268.
    Facility Operating License No. DPR-59: Amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: June 28, 2000 (65 FR 
39953).
    The supplemental information did not expand the scope of the 
application as originally noticed in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 9, 2000.
    No significant hazards consideration comments received: No.

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 application for amendments: February 7, 2000, as 
supplemented on August 9 and October 12, 2000.
    Brief description of amendments: The amendments modify the Salem 
Unit Nos. 1 and 2 Technical Specifications (TS), and revise 
surveillance requirements associated with Auxiliary Feedwater (AFW) 
Pump testing described in TS 4.7.1.2.b by replacing the current wording 
with that of improved Standard TSs, NUREG-1431, ``Standard Technical 
Specifications, Westinghouse Plants.''
    Date of issuance: December 5, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment Nos.: 238 and 219.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37428).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 5, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 8, 2000, as supplemented 
April 5, 2000, and October 25, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications through revision to the storage configuration 
requirements within the existing storage racks and taking credit for a 
limited amount of soluble boron.
    Date of issuance: December 7, 2000.
    Effective date: December 7, 2000.
    Amendment No.: 79.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17918).
    The April 5, 2000, and October 25, 2000, submittals provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 7, 2000.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant Units 1 and 2, Rhea County, Tennessee

    Date of application for amendment: March 10, 2000, as supplemented 
November 6 and 9, 2000 and November 21, 2000 (two letters).
    Brief description of amendment: Changed the Operating License to 
incorporate Physical Security/Contingency Plan--Tamper Indicating/Line 
Supervision Alarms Testing Frequency at Watts Bar Nuclear Plant (WBN) 
Units 1 and 2.

[[Page 81936]]

    Date of issuance: December 5, 2000.
    Effective date: December 5, 2000.
    Amendment No.: 29 and 29.
    Facility Operating License No. NPF-90: Amendment revises the 
Operating License.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56957). The November 6, 9, and 21, 2000, supplements provided 
clarifying information that did not change the scope of the initial 
proposed no significant hazards consideration determination.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 30, 2000, as 
supplemented November 15 and 22, 2000.
    Brief description of amendment: Allow a one-time-only increase in 
the diesel generator Action Completion Time from 72 hours to 10 days to 
facilitate repairs to an emergency diesel generator to improve 
reliability.
    Date of issuance: December 8, 2000.
    Effective date: December 8, 2000.
    Amendment No.: 30.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 2000 (65 FR 
66266). The November 15 and 22, 2000 supplements provided clarifying 
information that did not change the scope of the initial proposed no 
significant hazards consideration determimination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 8, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 10, 2000.
    Brief description of amendments: The amendments change Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report,'' to 
incorporate the latest, Nuclear Regulatory Commission (NRC)-approved 
methodology for analysis of large break loss-of-coolant accidents 
(LBLOCAs) for Comanche Peak Steam Electric Station, Units 1 and 2. The 
acceptability of this change to TS 5.6.5 is based upon the NRC staff's 
conclusion that the LBLOCA analysis methodology described in TXU 
Electric's Topical Report ERX-2000-002-P, ``Revised Large Break Loss of 
Coolant Accident Methodology,'' March 2000, is acceptable, as addressed 
in the associated Safety Evaluation.
    Date of issuance: October 6, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 80 and 80.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51363).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 6, 2000.
    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: May 2, 2000, as supplemented August 30, 
2000.
    Brief description of amendments: The amendments change the CPSES 
Security Plan to: (1) Allow response team members to perform 
compensatory measures for protective area intrusion detection or closed 
circuit television failure, and (2) to modify the patrol frequency for 
the protected area.
    Date of issuance: December 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 82 and 82.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Security Plan.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59226).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 5, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 21st day of December 2000.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-33012 Filed 12-26-00; 8:45 am]
BILLING CODE 7590-01-P