Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 51222-51236 [06-7137]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 167 (Tuesday, August 29, 2006)]
[Notices]
[Pages 51222-51236]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-7137]


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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 section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory

[[Page 51223]]

Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. The Act requires the Commission publish notice of any 
amendments issued, or proposed to be issued and 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 August 4, 2006 to August 17, 2006. The last 
biweekly notice was published on August 15, 2006 (71 FR 46929).

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. Within 60 days after the date of publication of this 
notice, 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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, 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 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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

[[Page 51224]]

limitations in the order granting leave to intervene, and have the 
opportunity to participate fully in the conduct of the hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: April 11, 2006.
    Description of amendment request: The proposed amendment would 
modify Technical Specification 5.6.5 ``Core Operating Limits Report 
(COLR)'' to add two U.S. Nuclear Regulatory Commission-approved topical 
reports to the COLR methodologies list. These topical reports allow the 
use of S-RELAP5 thermal-hydraulic analysis code for accident safety 
analyses.
    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 two topical reports have been reviewed and approved by the 
NRC for use in determining core operating limits. The core operating 
limits to be developed using the new methodologies for HBRSEP [H. B. 
Robinson Steam Electric Plant], Unit No. 2, will be established in 
accordance with the applicable limitations as documented in the NRC 
Safety Evaluation Reports. In a May 11, 2001, NRC Safety Evaluation 
Report, the NRC concluded that the S-RELAP5 code is capable of 
addressing the thermal-hydraulic response of the target non-LOCA 
[loss-of-coolant accident] events in a conservative manner and is, 
therefore, an acceptable replacement for the ANF-RELAP code. In the 
May 19, 2004, Safety Evaluation Report for Revision 1 to EMF-
2310(P)(A), the NRC concluded that the code remained acceptable for 
use for the non-LOCA events. In a March 15, 2001, Safety Evaluation 
Report, the NRC concluded that the code was acceptable for use for 
small break LOCA analyses at Westinghouse pressurized water 
reactors.
    The proposed change, by itself, does not impact the current 
design bases. The proposed change enables the use of new 
methodologies to re-analyze certain events. Revised analyses may 
either result in continued conformance with design bases, or may 
change the design bases. If design basis changes result from a 
revised analysis, then the specific design changes will be evaluated 
in accordance with HBRSEP, Unit No. 2, design change procedures and 
10 CFR 50.59.
    The proposed change does not involve physical changes to any 
plant structure, system, or component. Therefore, the probability of 
occurrence for a previously analyzed accident is not significantly 
increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fission product barriers during the analyzed accident, the 
availability and successful functioning of the equipment assumed to 
operate in response to the analyzed event, and the setpoints at 
which these actions are initiated. The proposed methodologies will 
ensure that the plant continues to meet applicable design and safety 
analyses acceptance criteria. The proposed change does not affect 
the performance of any equipment used to mitigate the consequences 
of an analyzed accident. As a result, no analysis assumptions are 
impacted and there are no adverse effects on the factors that 
contribute to offsite or onsite dose as a result of an accident. The 
proposed change does not affect setpoints that initiate protective 
or mitigative actions. The proposed change ensures that plant 
structures, systems, and components are maintained consistent with 
the safety analysis and licensing bases. Based on this evaluation, 
there is no significant increase in the consequences of a previously 
analyzed accident.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated.
    The proposed change does not involve any physical alteration of 
plant systems, structures, or components, other than allowing for 
fuel design in accordance with NRC-approved methodologies. No new or 
different equipment is being installed. No installed equipment is 
being operated in a different manner. There is no change to the 
parameters within which the plant is normally operated or in the 
setpoints that initiate protective or mitigative actions. As a 
result, no new failure modes are being introduced. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety.
    There is no impact on any margin of safety resulting from the 
incorporation of these new topical reports into the Technical

[[Page 51225]]

Specifications. If design basis changes result from a revised 
analysis that uses these new methodologies, the specific design 
changes will be evaluated in accordance with HBRSEP, Unit No. 2, 
design change procedures and 10 CFR 50.59. Any potential reduction 
in the margin of safety would be evaluated for that specific design 
change.
    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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: July 17, 2006.
    Description of amendment request: The proposed amendment would 
revise the containment pressure requirements specified in Surveillance 
Requirements 3.6.8 and 5.5.16 due to a revision in the Loss-of-Coolant 
Accident (LOCA) containment pressure analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The revised post-LOCA containment pressure and 
temperature analysis used more conservative assumptions and the 
increase in the calculated peak pressure was approximately 1 psig. 
The revised value of 41.49 psig remains less than the containment 
design pressure of 42 psig. The increase in the calculated peak 
temperature was approximately 2 [deg]F, which was analyzed to have 
no impact on structures or equipment. Although there is an increase 
in the calculated pressure, the allowable containment leakage rate, 
as measured at the peak pressure, is not being changed. Since there 
is no increase in the allowable leakage, there is no increase in 
consequences. The proposed change is related to containment pressure 
analysis. There are no physical changes being made to the plant, or 
to the manner in which the plant is operated. Surveillance 
procedures for containment leakage have been conservatively testing 
at pressures in excess of 42 psig and surveillance procedures for 
the Isolation Valve Seal Water System have been conservatively 
testing at pressures in excess of 46.2 psig. The change can have no 
impact on the probability of an accident occurring. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated. There 
are no physical changes being made to the plant or to the manner in 
which the plant is operated. Surveillance procedures for containment 
leakage have been conservatively testing at pressures in excess of 
42 psig and surveillance procedures for the Isolation Valve Seal 
Water System have been conservatively testing at pressures in excess 
of 46.2 psig. The revised containment analysis results in a 
calculated peak containment pressure that remains less than the 
containment design pressure. The increase in the calculated peak 
temperature was analyzed to have no impact on structures or 
equipment. Therefore, this change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    No. The proposed change does not involve a significant reduction 
in the margin of safety. The proposed change imposes more 
conservative surveillance test requirements. The calculated increase 
in post-LOCA peak containment pressure is only 1 psig and the 
revised value of 41.49 psig remains less than the containment design 
pressure of 42 psig. The increase in the calculated peak temperature 
was approximately 2 [deg]F, which was analyzed to have no impact on 
structures or equipment. Although there was an increase in the 
calculated pressure, the allowable containment leakage rate, as 
measured at the peak pressure, is not being changed. Therefore, this 
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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: July 12, 2006.
    Description of amendment request: The proposed amendment would 
modify Conditions, Required Actions and Completion Times associated 
with the inoperability of one or more emergency diesel generators 
(EDGs) in Technical Specification (TS) 3.8.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This license amendment request proposes a change to extend the 
Technical Specification 3.8.1, ``AC Sources-Operating,'' Completion 
Time. This change allows a single EDG to be inoperable for 7 days 
more than Technical Specification 3.8.1 currently provides. The 
Required Actions for CTG [Combustion Turbine Generator] 11-1 are 
also removed from Condition A and TSTF [Technical Specification Task 
Force] -439 is implemented for TS 3.8.1, removing the second 
Completion Times.
    The EDGs are safety related components which provide backup 
electrical power supply to the onsite ESF [Engineered Safety 
Feature] power distribution system. CTG 11-1 provides backup 
electrical power to the Division 1 power distribution system. 
Neither the EDGs nor CTG 11-1 are accident initiators, thus these 
changes do not increase the probability of a previously evaluated 
accident.
    The plant ESF power distribution systems consist of two 
divisions for 100% redundancy. Accident analyses demonstrate that 
only one division is required for accident mitigation. Thus, with 
one division inoperable the other division is capable of performing 
the required safety function. Design basis analyses are not required 
to be performed assuming extended loss of all power supplies to the 
plant ESF power distribution system. Thus, this change does not 
involve a significant increase in the consequences of a previously 
analyzed accident.
    The proposed change also eliminates the second Completion Time 
from TS 3.8.1. These second Completion Times are not an initiator to 
any accident previously evaluated. As a result, the probability of 
an accident previously evaluated is not affected. The consequences 
of an accident during the revised Completion Times are no different 
than the consequences of the same accident during the existing 
Completion Times. As a result, the consequences of an accident 
previously evaluated are not affected by this change. Therefore, the 
proposed change does not involve a significant increase in the

[[Page 51226]]

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 changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The changes 
do not alter any assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    This license amendment request proposes Technical Specification 
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable EDG to 14 days. These 
changes allow an emergency diesel generator to be inoperable for 7 
days more than TS 3.8.1 currently provides.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed TS changes on the availability of an 
electrical power supply to the plant emergency safeguards features 
systems. These assessments concluded that the proposed TS changes do 
not involve a significant increase in the risk of power supply 
unavailability.
    This license amendment request proposes TS changes to remove the 
Required Actions for CTG 11-1 from TS 3.8.1 Condition A. If CTG 11-1 
is inoperable at the same time that any single EDG is inoperable for 
the entire proposed 14 day period with no other equipment in 
maintenance, the risk remains within RG [Regulatory Guide] 1.174 
(Reference 2 [in the application]) thresholds for a ``very small'' 
classification.
    The proposed change to delete the second Completion Time does 
not alter the manner in which safety limits, limiting safety system 
settings, or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed changes will not result in plant operation in a 
configuration outside of the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Acting Branch Chief: Martin Murphy.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, (MPS2) New London County, Connecticut

    Date of amendment request: June 13, 2006.
    Description of amendment request: The proposed amendment would 
allow changes to the Technical Specifications based on the radiological 
dose analysis margins obtained by using an alternate source term 
consistent with 10 CFR 50.67.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No.
    The proposed amendment does not involve a significant increase 
in the probability or consequence of an accident previously 
analyzed. The MPS2 control room emergency ventilation system only 
functions following the initiation of a design basis radiological 
accident. Therefore, the change to the value used for in-leakage 
rate test acceptance criteria following a design basis accident will 
not increase the probability of any previously analyzed accident. 
The proposed 200 cfm control room habitability envelope inleakage 
surveillance acceptance criteria has no adverse impact on control 
room habitability analyses for postulated toxic chemical release 
events. These habitability analyses do not credit automatic or 
manual isolation of the control room fresh air ventilation flow 
during a toxic chemical release event. The control room's forced 
ventilation fresh air exchange rate (e.g., 800 cfm) is much greater 
than the proposed 200 cfm envelope inleakage rate acceptance 
criteria. The MPS2 containment purge valve isolation signal is not 
credited in the accident analyses. The requirements contained in 
this specification do not meet any of 10 CFR 50.36(c)(2)(ii) 
criteria on items for which technical specifications must be 
established. Deletion of this technical specification will not 
increase the probability of any previously analyzed accident. The 
MPS2 containment and the containment systems function to prevent or 
control the release of radioactive fission products following a 
postulated accident. Therefore, the change to the value used for 
primary to secondary leak rate acceptance criteria, and for all 
secondary containment bypass leakage paths following a design basis 
accident, will not increase the probability of any previously 
analyzed accident.
    These systems are not initiators of any design bases accident. 
Revised dose calculations, which take into account the changes 
proposed by this amendment and the use of the alternative source 
term, have been performed for the MPS2 design basis radiological 
accidents. The results of these revised calculations indicate that 
public and control room doses will not exceed the limits specified 
in 10 CFR 50.67 and Regulatory Guide 1.183. There is not a 
significant increase in predicted dose consequences for any of the 
analyzed accidents. Therefore, the proposed changes do not involve a 
significant increase in the consequences of any previously analyzed 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    No.
    The implementation of the proposed changes does not create the 
possibility of an accident of a different type than was previously 
evaluated in the UFSAR. Although the proposed changes could affect 
the operation of the control room emergency ventilation system, and 
the containment and containment systems following a design basis 
radiological accident, none of these changes can initiate a new or 
different kind of accident since they are only related to system 
capabilities that provide protection from accidents that have 
already occurred. These changes do not alter the nature of events 
postulated in the UFSAR nor do they introduce any unique precursor 
mechanisms. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from those 
previously analyzed.
    3. Involve a significant reduction in the margin of safety?
    No.
    The implementation of the proposed changes does not reduce the 
margin of safety. The proposed changes for the control room 
ventilation system, and the containment and containment systems do 
not affect the ability of these systems to perform their intended 
safety functions to maintain dose less than the required limits 
during design basis radiological events. The radiological analysis 
results, when compared with the revised TEDE [total effective dose 
equivalent] acceptance criteria, meet the applicable limits. These 
acceptance criteria have been developed for application to analyses 
performed with alternative source terms. These acceptance criteria 
have been developed for the purpose of use in design basis accident 
analyses such that meeting the stated limits demonstrates adequate 
protection of public health and safety. It is thus concluded that 
the margin of safety will not be reduced by the implementation of 
the changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Acting Branch Chief: Brooke D. Poole.

[[Page 51227]]

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: July 19, 2006.
    Description of amendment request: The proposed amendment will 
revise reactor core safety limits Technical Specifications (TSs) and 
relocate the reactor core safety limit figure to the Core Operating 
Limits Report (COLR) in the Millstone Power Station, Unit 3, Technical 
Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1:
    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The relocation of cycle-specific core operating limits from the 
TS to the COLR and the addition of the RCS [reactor coolant system] 
total flow rate to the COLR has no influence or impact on the 
probability or consequences of a design basis accident. Adherence to 
the COLR and methodologies acceptable for establishing COLR 
parameters continues to be controlled by TS. The proposed amendment 
still requires exactly the same actions to be taken when or if 
limits are exceeded. Each accident analysis addressed in the final 
safety analysis report (FSAR) will be examined with respect to 
changes in cycle-dependent parameters, which are obtained from 
application of the NRC-approved reload design methodologies, to 
ensure that the transient evaluation of new core designs are bounded 
by previously accepted analyses. This examination, which will be 
performed in accordance with the requirements of 10 CFR 50.59, 
ensures that future core designs will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The relocation and addition of the cycle-specific variables to 
the COLR and adding new document references in TS Section 6.9.1.6.b 
does not influence or impact, nor does it contribute in any way to 
the probability or consequences of an accident. No safety-related 
equipment, safety function, or plant operations will be altered as a 
result of these proposed changes. The cycle-specific variables are 
calculated using NRC-approved methods and submitted to the NRC to 
allow the staff to continue to trend the values of these limits. The 
TS will continue to require operation within the required core 
operating limits and appropriate actions will be taken when or if 
limits are exceeded. Therefore the proposed amendment does not in 
any way create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes have no impact on plant equipment 
operation. The proposed changes do not revise any setpoints or 
acceptance criteria assumed in the accident analyses. Therefore, the 
proposed changes will not result in 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Acting Branch Chief: Brooke D. Poole.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2 (IP2), Westchester County, New York

    Date of amendment request: July 10, 2006.
    Description of amendment request: Energy Nuclear Operations, Inc., 
is planning to operate an Independent Spent Fuel Storage Installation 
(ISFSI) facility at IP2 using the HOLTEC HI-STORM 100 Cask System. To 
support this activity, the proposed amendment adds Spent Fuel Cask 
loading requirements to IP2 Technical Specifications (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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise the Indian Point Unit 2 TSs 
associated with the SFP [spent fuel pool] to assure that the 
regulatory requirements related to criticality in the SFP and 
applied to the Holtec HI-STORM 100 Multi-Purpose Canister MPC-32 
when in the SFP are reflected in the IP2 TS. The proposed change 
does not require any physical changes to Part 50 structures, 
systems, or components, nor will their performance requirements be 
altered. Therefore, the response of the plant to previously analyzed 
accidents and related radiological releases will not be adversely 
impacted, and will bound those postulated during cask loading 
activities in the cask storage area.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Existing fuel handling procedures and associated administrative 
controls remain applicable for cask loading operations within the 
SFP. Additionally, the soluble boron concentration required to 
maintain keff <= 0.95 for postulated criticality accidents 
associated with cask loading operations was also evaluated. The 
results of the analyses, using a methodology previously approved by 
the NRC [Nuclear Regulatory Commission], demonstrate that the amount 
of soluble boron required to compensate for the positive reactivity 
associated with these postulated accidents (371 ppm [parts per 
million]) remains well below the existing spent fuel pit minimum 
boron concentration limit of 2000 ppm. Accordingly, the same limit 
has been proposed for cask loading operations in the cask storage 
area.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    An NRC approved methodology was used to perform the criticality 
analysis which provides the basis to incorporate a new family of 
burnup versus enrichment curves, for various cooling times, into the 
plant Technical Specifications to ensure criticality requirements 
are met during spent fuel cask loading. Accordingly, the existing 
minimum boron concentration limit for the spent fuel pit of 2000 ppm 
will continue to remain bounding during cask loading operations. 
This determination accounts for uncertainties at a 95 percent 
probability, 95 percent confidence level. Should it be postulated 
that a boron dilution event does occur during this time period, keff 
will remain less than 1.0 should the cask storage area become fully 
flooded with unborated water.

    Therefore, there will not be a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Richard J. Laufer.

[[Page 51228]]

Exelon Generation Company, LLC, Docket No. 50-249, Dresden Nuclear 
Power Station (DNPS), Unit 3, Grundy County, Illinois

    Date of amendment request: July 21, 2006.
    Description of amendment request: The proposed amendment would 
revise the values of the safety limit minimum critical power ratio 
(SLMCPR) in Technical Specification Section 2.1.1, ``Reactor Core SLs 
[Safety Limits].'' Specifically, the proposed change would require that 
for Unit 3, the minimum critical power ratio (MCPR) for Global Nuclear 
Fuel fuel shall be >= 1.10 for two recirculation loop operation, or >= 
1.11 for single recirculation loop operation. Additionally, the 
proposed change would require that MCPR for Westinghouse fuel shall be 
>= 1.12 for two recirculation loop operation, or >= 1.14 for single 
recirculation loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC 
[Nuclear Regulatory Commission]-approved methods to ensure that fuel 
performance during normal, transient, and accident conditions is 
acceptable. The proposed change conservatively establishes the 
SLMCPR for DNPS, Unit 3, Cycle 20 such that the fuel is protected 
during normal operation and during plant transients or anticipated 
operational occurrences (AOOs).
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during plant transients or AOOs. 
Operational limits will be established based on the proposed SLMCPR 
to ensure that the SLMCPR is not violated. This will ensure that the 
fuel design safety criterion (i.e., that at least 99.9% of the fuel 
rods do not experience transition boiling during normal operation 
and AOOs) is met. Since the proposed change does not affect 
operability of plant systems designed to mitigate any consequences 
of accidents, the consequences of an accident previously evaluated 
are not expected to increase.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed change does not involve any plant configuration 
modifications or changes to allowable modes of operation. The 
proposed change to the SLMCPR assures that safety criteria are 
maintained for DNPS, Unit 3, Cycle 20.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SLMCPR provides a margin of safety by ensuring that at least 
99.9% of the fuel rods do not experience transition boiling during 
normal operation and AOOs if the MCPR limit is not violated. The 
proposed change will ensure the current level of fuel protection is 
maintained by continuing to ensure that at least 99.9% of the fuel 
rods do not experience transition boiling during normal operation 
and AOOs if the MCPR limit is not violated. The proposed SLMCPR 
values were developed using NRC-approved methods. Additionally, 
operational limits will be established based on the proposed SLMCPR 
to ensure that the SLMCPR is not violated. This will ensure that the 
fuel design safety criterion (i.e., that no more than 0.1% of the 
rods are expected to be in boiling transition if the MCPR limit is 
not violated) is met.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio

    Date of amendment request: June 6, 2006.
    Description of amendment request: The proposed amendment would 
revise the Ventilation Filter Test Program (VFTP) in Technical 
Specification (TS) 5.5.7. The license amendment is a corrective action 
to revise the flow rate units specified in the VFTP from standard cubic 
feet per minute to cubic feet per minute. This amendment will ensure 
the PNPP TS are consistent with plant design documentation, testing 
criteria, and the industry.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response:
    The ESF [Engineered Safety Feature] Ventilation systems reduce 
the concentration of airborne radioactive contaminants following a 
design basis accident and therefore are not initiators of design 
bases accidents. The proposed amendment does not change the manner 
in which the ESF ventilation systems are operated or tested. 
Implementation of the proposed amendment will ensure the ESF 
ventilation systems perform their function when called upon and does 
not affect the plant operations, design function or analysis that 
verifies the capability of a [plant] structures, systems or 
components.
    The proposed amendment does not affect the design of the ESF 
ventilation systems, the operational characteristics of the ESF 
ventilation systems, the interfaces between the ESF ventilation 
systems and those plant systems they support, or the reliability of 
the ESF ventilation systems.
    Therefore, the ESF ventilation systems will be capable of 
performing their accident mitigation function and there is no 
increase in the probability or consequences of an accident already 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed amendment introduces no new mode of plant operation 
and does not involve a physical modification to the plant. New 
equipment is not installed with the proposed amendment, nor does the 
proposed amendment cause existing equipment to be operated in a new 
or different manner.
    Since the proposed changes do not involve a change to the plant 
design or operation, no new system interactions are created by this 
change. The proposed amendment does not produce any parameters or 
conditions that

[[Page 51229]]

could contribute to the initiation of accidents different from those 
already evaluated in the Updated Safety Analysis Report.
    The changes to the VFTP do not affect the assumed accident 
performance of the ESF Ventilation systems, nor [sic] any plant 
structure, system or component previously evaluated.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed amendment does not impact the ESF ventilation 
systems performance, including the capability for each ESF 
ventilation system to attain and maintain required air flow assumed 
in the plant safety analysis.
    The proposed amendment does not involve a significant reduction 
in a margin of safety since the operability of the ESF ventilations 
[sic] systems continues to be determined as required to support the 
capability of the ESF ventilations [sic] systems to provide the 
required ventilation, filtration and temperature control to mitigate 
the consequences of an accident.

    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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Daniel S. Collins.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: May 25, 2006.
    Description of amendment request: The proposed license amendment 
revises the requirements in the Crystal River Unit 3 Improved Technical 
Specification related to steam generator tube integrity. The licensee 
states that the changes are consistent with NRC-approved Technical 
Specification (TS) Task Force (TSTF) Change Traveler, TSTF-449, ``Steam 
Generator Tube Integrity,'' Revision 4. The availability of this 
technical specification improvement was announced in the Federal 
Register on May 6, 2005, as part of the consolidated line item 
improvement process.
    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 proposed change requires a SG [Steam Generator] Program that 
includes performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown and all anticipated transients included 
in the design specification). The SG performance criteria are based 
on tube structural integrity, accident induced leakage, and 
operational LEAKAGE.
    A SGTR [steam generarator tube rupture] event is one of the 
design basis accidents that are analyzed as part of a plant's 
licensing basis. In the analysis of a SGTR event, a bounding primary 
to secondary LEAKAGE rate equal to the operational LEAKAGE rate 
limits in the licensing basis plus the LEAKAGE rate associated with 
a double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB [main steamline 
break], rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by NEI [Nuclear Energy Institute] 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates 
a balance of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT 1-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT 1-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT 1-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated

    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

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

    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the

[[Page 51230]]

tube integrity by implementing the SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-266, Point Beach Nuclear 
Plant (PBNP), Unit 1, Town of Two Creeks, Manitowoc County, Wisconsin

    Date of amendment request: July 11, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.8, ``Steam Generator (SG) 
Program,'' to exclude the portion of the tube below 17 inches from the 
top of the tubesheet from the SG tube inspection requirements for Unit 
1 on a one-time basis for a single operating cycle. In addition, 
administrative changes are proposed to correct a page number in the TS 
table of contents and delete two blank pages in TS Section 5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration which 
is presented below.

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The proposed change revises Technical Specification (TS) 5.5.8, 
``Steam Generator (SG) Program'' to redefine the PBNP Unit 1 primary 
pressure boundary for purposes of the SG tube inspection 
requirements on a one-time basis for Unit 1 Refueling Outage 30 and 
the subsequent operating cycle. The redefined primary pressure 
boundary is relocated from the seal weld at the bottom of the SG 
tube to the tube-to-tubesheet mechanical interface. The required 
structural integrity margins of the SG tubes in this area are 
unaffected by this change and will be maintained by the SG 
tubesheet. SG tubes are hydraulically expanded into the tubesheet. 
Steam generator tube rupture is constrained by the tubesheet for 
tubes with cracks in the tubesheet. This constraint results from the 
hydraulic expansion process which restricts further expansion of the 
tube, thermal expansion mismatch between the tube and tubesheet and 
from the differential pressure between the primary and secondary 
side. Thermal expansion and differential pressure also restrain the 
tube axially. For conservatism, hydraulic preload was not factored 
into the analysis.
    The proposed change continues to require that the SG Program 
include performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown and all anticipated transients included 
in the design specification).
    The SG performance criteria are based on tube structural 
integrity, accident induced leakage, and operational LEAKAGE. The 
analysis shows that structural integrity retains acceptable safety 
factors against burst under normal steady state full power operation 
primary-to-secondary pressure differential and against burst applied 
to the design basis accident primary-to-secondary pressure 
differentials. The analysis also shows that accident induced leakage 
is bound by twice the normal operating leakage and well below the 
accident analysis assumption for each stream generator. The primary 
to secondary operational LEAKAGE limit is not changed.
    The planned inspection and supporting analysis provide 
reasonable assurance that the SG tubing will remain capable of 
fulfilling its specific safety function of maintaining reactor 
coolant pressure boundary integrity throughout the operating cycle 
and in the unlikely event of a design basis accident. The proposed 
change does not, therefore, significantly increase the probability 
of an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
The plant technical specification limits for operational LEAKAGE and 
for DOSE EQUIVALENT I-131 in primary coolant, which ensure the plant 
is operated within its analyzed condition, are unaffected by the 
proposed change. Therefore, the proposed change does not 
significantly increase the consequences of any accident previously 
evaluated.
    The proposed change does not significantly affect the 
probability of any event initiators. There will be no change to 
normal plant operating parameters, engineered safety feature 
actuation setpoints, accident mitigation capabilities, or accident 
analysis assumptions or inputs.
    Therefore, the probability or consequences of any accident 
previously evaluated will not be significantly increased as a result 
of the proposed change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    Implementation of the proposed change will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. Primary to secondary 
leakage that may be experienced during all plant conditions will 
continue to be monitored to ensure it remains within current 
accident analysis assumptions. The proposed change does not affect 
the method of operation of the SGs, or primary or secondary coolant 
chemistry controls. In addition, the proposed change does not impact 
any other plant system or component.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. Equipment important to safety will continue 
to operate as designed. The changes do not result in any event 
previously deemed incredible being made credible. The changes do not 
result in adverse conditions or result in any increase in the 
challenges to safety systems. Therefore, the proposed change does 
not create the possibility of a new or different type of accident 
from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The steam generators (SGs) are an integral part of the reactor 
coolant pressure boundary and, as such, are relied upon to maintain 
the primar