Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 17944-17959 [E7-6632]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 68 (Tuesday, April 10, 2007)]
[Notices]
[Pages 17944-17959]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-6632]


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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 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 March 16, 2007 to March 29, 2007. The last 
biweekly notice was published on March 27, 2007 (72 FR 14303).

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, Rulemaking, 
Directives and Editing 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

[[Page 17945]]

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 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.

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

    Date of amendment request: November 13, 2006.
    Description of amendment request: The proposed amendment changes 
the technical specification (TS) testing frequency for the surveillance 
requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' The proposed 
change revises the test frequency of SR 3.1.4.2, control rod scram time 
testing, from ``120 days cumulative operation in MODE 1'' to ``200 days 
cumulative operation in Mode 1.''
    AmerGen has reviewed the proposed no significant hazards 
consideration determination published in the Federal Register on August 
23, 2004 (69 FR 51864), as part of the consolidated line item 
improvement process (CLIIP) and has concluded that the proposed 
determination presented in the notice is applicable to Clinton Power 
Station, Unit No. 1.
    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.

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. 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 proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    Based on the above, the proposed change presents no significant 
hazards consideration under the standards set forth in 10 CFR 50.92(c), 
and accordingly, a finding of ``no significant hazards consideration'' 
is justified.
    Therefore, the NRC staff proposes to determine that the amendment 
request involves 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: Russell Gibbs.

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

    Date of amendment request: January 26, 2007.
    Description of amendment request: The proposed amendment would 
revise technical specifications (TS) requirements for unavailable 
barriers by

[[Page 17946]]

adding limiting condition for operation (LCO) 3.0.9. This would 
establish conditions under which TS systems would remain operable when 
required physical barriers are not capable of providing their related 
support function. Also, the proposed amendment would make editorial 
changes to LCO 3.0.8 to be consistent with the terminology in LCO 
3.0.9.
    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 by a reference to a generic analysis published in the 
Federal Register on October 3, 2006 (71 FR 58444), which is presented 
below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. 
The postulated initiating events which may require a functional 
barrier are limited to those with low frequencies of occurrence, and 
the overall TS system safety function would still be available for 
the majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on the 
allowance provided by proposed LCO 3.0.9 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.9. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns.
    Therefore, this 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.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those 
with low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG 1.177. A 
bounding risk assessment was performed to justify the proposed TS 
changes. This application of LCO 3.0.9 is predicated upon the 
licensee's performance of a risk assessment and the management of 
plant risk. The net change to the margin of safety is insignificant 
as indicated by the anticipated low levels of associated risk (ICCDP 
and ICLERP) as shown in Table 1 of Section 3.1.1 in the Safety 
Evaluation.
    Therefore, this 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 Branch Chief: L. Raghavan.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: January 18, 2007.
    Description of amendment request: The proposed amendment would 
revise the expiration limit for the reactor coolant system Pressure/
Temperature (P/T) limit graphs in Technical Specifications (TS); revise 
the adjusted reference temperature for the reactor vessel; and revise 
the Low Temperature Overpressure Protection (LTOP) arming temperature 
value specified in TSs. It would also make editorial changes in the use 
of inequality signs in TSs associated with the LTOP arming temperature 
in order to make them consistent.
    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 does not affect the accident initiators or 
mitigation assumptions associated with any of the accidents 
previously evaluated. Operating restrictions on pressure-temperature 
conditions for the reactor pressure vessel provide assurance that 
reactor vessel integrity will be maintained under accident or 
transient conditions. The proposed change uses approved criteria and 
analysis methods to update the time period for which the current 
operating limits remain valid.
    The LTOP system performs an automatic function by opening relief 
valves if reactor coolant system pressure reaches a temperature-
dependent limit. The proposed change includes establishing a more 
restrictive temperature limit for when this system must be in 
service, to reflect the material condition of the reactor vessel at 
the new EFPY limit proposed for the pressure-temperature graphs. The 
mitigation function and capability of the LTOP system is not being 
changed by this request.
    Therefore, the proposed changes do 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
    There are no new accident initiators being introduced by this 
proposed change. The proposed change does not involve installation 
of new plant equipment, modification of existing equipment, or 
changes in the way that plant equipment is operated. Pressure-
temperature operating limits depicted by graphs in the technical 
specifications will not be changed and will continue to be used by 
plant operators. A change in the LTOP system arming temperature will 
assure that the graphs remain valid for the proposed new operating 
period of 27.2 EFPY [effective full power years].
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    Operating limits on pressure and temperature conditions for the 
reactor coolant system (RCS) are important to assure that the RCS 
pressure boundary stresses are within analyzed limits. Margins of 
safety are inherent in the analysis methods, assumptions, and limits 
specified in regulations and guidance documents. The proposed change 
is based on NRC-accepted methods, assumptions and limits and 
maintains the required margin of safety.
    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

[[Page 17947]]

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 Acting Branch Chief: Douglas V. Pickett.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 (IP2) and 3 (IP3), Westchester 
County, New York

    Date of amendment request: March 13, 2007.
    Description of amendment request: The amendment would revise 
License Condition 2.K for IP2 and License Condition 2.H for IP3, which 
require the implementation and maintenance of an approved Fire 
Protection Program for each unit.
    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 changes are strictly an administrative relocation 
of the specific fire protection SER [safety evaluation report] 
references and do not modify any requirements of the fire protection 
programs.
    Therefore, the proposed changes do 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.
    The proposed changes are strictly an administrative relocation 
of the specific fire protection SER references and do not modify any 
requirements of the fire protection programs.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are strictly an administrative relocation 
of the specific fire protection SER references and do not modify any 
requirements of the fire protection programs.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Douglas V. Pickett.

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

    Date of amendment request: March 22, 2007.
    Description of amendment request: The proposed amendment will 
revise the test acceptance criteria specified in Technical 
Specification Surveillance Requirement (SR) 3.8.1.10 for the diesel 
generator endurance test.
    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?
    No. The proposed change revises the acceptance criteria to be 
applied to an existing surveillance test of the facility emergency 
diesel generators (DGs). Performing a surveillance test is not an 
accident initiator and does not increase the probability of an 
accident occurring. The proposed new acceptance criteria will assure 
that the DGs are capable of carrying the peak electrical loading 
assumed in the various existing safety analyses which take credit 
for the operation of the DGs. Establishing acceptance criteria that 
bound existing analyses validates the related assumption used in 
those analyses regarding the capability of equipment to mitigate 
accident conditions. Therefore the proposed change does 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?
    No. The proposed change revises the test acceptance criteria for 
a specific performance test conducted on the existing DGs. The 
proposed change does not involve installation of new equipment or 
modification of existing equipment, so no new equipment failure 
modes are introduced. The proposed revision to the DG surveillance 
test acceptance criteria also is not a change to the way that the 
equipment or facility is operated and no new accident initiators are 
created. Therefore the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The conduct of performance tests on safety-related plant 
equipment is a means of assuring that the equipment is capable of 
maintaining the margin of safety established in the safety analyses 
for the facility. The proposed change in the DG technical 
specification surveillance test acceptance criteria is consistent 
with values assumed in existing safety analyses and is consistent 
with the design rating of the DGs. Therefore the proposed change 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Douglas V. Pickett.

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

    Date of amendment request: February 15, 2007.
    Description of amendment request: The proposed changes would revise 
Technical Specification (TS) 3.10.1 to expand its scope to include 
provisions for reactor coolant temperature excursions greater than 212 
[deg]F as a consequence of inservice leak and hydrostatic testing, and 
as a consequence of scram time testing initiated in conjunction with an 
inservice leak or hydrostatic test, while considering operational 
conditions to be in Mode 4, which is defined to be reactor coolant 
temperature less than or equal to 212 [deg]F.
    This change was proposed by the industry's TS Task Force (TSTF) and 
is designated TSTF-484. The NRC staff issued a notice of opportunity 
for comment in the Federal Register on August 21, 2006 (71 FR 48561), 
on possible amendments concerning TSTF-484, including a model safety 
evaluation and model no significant hazards (NSHC) determination, using 
the consolidated line item improvement process (CLIIP). The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 27,

[[Page 17948]]

2006 (71 FR 63050). The licensee affirmed the applicability of the 
following NSHC determination in its application dated February 15, 
2007.
    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.
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or consequences of an accident 
previously evaluated. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Criterion 2: The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. 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. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3: The proposed change does not involve a significant 
reduction in a margin of safety.
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    Based on the above, the NRC staff concludes that the proposed 
change presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 
no significant hazards consideration is justified.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Douglas V. Pickett.

Exelon Generation Company, LLC (EGC), Docket Nos. 50-373 and 50-374, 
LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: November 17, 2006.
    Description of amendment request: The proposed amendments would 
replace references to Section XI of the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code (Code) with a 
reference to the ASME Code of Operation and Maintenance of Nuclear 
Power Plants (OM Code) in Technical Specification (TS) 5.5.7, 
``Inservice Testing Program [IST].'' These proposed changes are 
consistent with the implementation of the LSCS, Units 1 and 2 third 10-
year IST program in accordance with the requirements of Title 10 of the 
Code of Federal Regulations (10 CFR) Section 50.55a, ``Codes and 
standards,'' paragraph (f), ``Inservice testing requirements.'' The 
third 10-year interval for LSCS, Units 1 and 2 is scheduled to start on 
October 12, 2007.
    In addition to the replacement of the references, EGC is also 
adding provisions in TS 5.5.7, item b, to only apply Surveillance 
Requirement (SR) 3.0.2 to those inservice testing frequencies of two 
years or less. These proposed changes are based on TS Task Force (TSTF) 
Traveler No. 479-A (TSTF-479-A), Revision 0, ``Changes to Reflect 
Revision of 10 CFR 50.55a,'' as modified by TSTF-497, Revision 0, 
``Limit Inservice Testing Program SR 3.0.2 Application to Frequencies 
of 2 Years or Less'' and approved by the NRC in December 6, 2005, and 
October 4, 2006.
    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?
    Response: No.
    The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to 
conform to the requirements of 10 CFR 50.55a, ``Codes and 
standards,'' paragraph (f) regarding the inservice testing of pumps 
and calves for the Third 10-year Interval. The current TS reference 
the [American Society of Mechanical Engineers] ASME Boiler and 
Pressure Vessel Code, Section XI, requirements for the inservice 
testing of ASME Code Class 1, 2, and 3 pumps and valves. The 
proposed changes would reference the ASME OM Code, which is 
consistent with 10 CFR 50.55a, paragraph (f), ``Inservice testing 
requirements,'' and approved for use by the NRC. In addition, 
provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only 
applied to those inservice testing frequencies of two years or less. 
The definitions of the frequencies are not changed by this license 
amendment request.
    The proposed changes are administrative in nature, do not affect 
any accident initiators, do not affect the ability of LSCS to 
successfully respond to previously evaluated accidents and do not 
affect radiological assumptions used in the evaluations. Thus, the 
radiological consequences of any accident previously evaluated are 
not increased.
    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?
    Response: No.
    The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to 
conform to the requirements of 10 CFR 50.55a(f) regarding the 
inservice testing of pumps and valves for the Third 10-year 
Interval. The current TS reference the ASME Boiler and Pressure 
Vessel Code, Section XI, requirements for the inservice testing of 
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes 
would reference the ASME OM Code, which is consistent with the 10 
CFR 50.55a(f) and approved for use by the NRC. In addition, 
provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only 
applied to those inservice testing frequencies of two years or less. 
The definitions of the frequencies are not changed by this license 
amendment request.
    The proposed changes to TS Section 5.5.7 do not affect the 
performance of any LSCS structure, system, or component credited 
with mitigating any accident previously evaluated and do not 
introduce any new modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to 
conform to the requirements of 10 CFR 50.55a(f) regarding the 
inservice testing of pumps and valves for the Third 10-Year 
Interval. The current TS reference the ASME Boiler and Pressure 
Vessel Code, Section XI, requirements for the inservice testing of 
ASME Code Class 1, 2,

[[Page 17949]]

and 3 pumps and valves. The proposed changes would reference the 
ASME OM Code, which is consistent with the 10 CFR 50.55a(f) and 
approved for use by the NRC. In addition, provisions modifying TS 
5.5.7, item b, clarify that SR 3.0.2 is only applied to those 
inservice testing frequencies of two years or less. The definitions 
of the frequencies are not changed by this license amendment 
request.
    The proposed changes do not modify the safety limits setpoints 
at which protective actions are initiated and do not change the 
requirements governing operation or availability of safety equipment 
assumed to operate to preserve the margin of safety.
    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 
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: Russell Gibbs.

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

    Date of amendment request: June 2, 2006.
    Description of amendment request: The proposed amendments 
incorporates revised 10 CFR Part 20 requirements for Limerick 
Generating Station Units 1 and 2 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.
    Updating the Technical Specifications (TS) to be consistent with 
10 CFR Part 20 has no impact on plant structures, systems, or 
components, does not affect any accident initiators, and does not 
change any safety analysis. Therefore, the proposed changes do not 
involve an 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.
    Updating the TS to be consistent with 10 CFR Part 20 will not 
change any equipment, require new equipment to be installed, or 
change the way current equipment operates. No credible new failure 
mechanisms, malfunctions, or accident initiators are created by the 
proposed changes.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Updating the TS to be consistent with 10 CFR Part 20 does not 
adversely affect existing plant safety margins or the reliability of 
equipment assumed to operate in the safety analysis. As such, there 
are no changes being made to safety analysis assumptions, safety 
limits or limiting safety system settings that would adversely 
affect plant safety as a result of the proposed changes. Therefore, 
the proposed changes do not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: February 12, 2007.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification (TS) Limiting Condition for 
Operation 3.9.4, ``Containment Penetrations'', to allow penetrations 
included under TS 3.9.4(c) to be opened during core alterations or 
movement of irradiated fuel, under administrative controls. This change 
is based on the TS Task Force Traveler No. 312-A, Revision 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. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow containment penetrations 
identified under Technical Specification 3.9.4(c) to remain open 
during fuel movement and core alterations. These penetrations are 
normally closed during this time period to prevent the release of 
radioactive material in the event of a Fuel Handling Accident inside 
containment. These penetrations are not initiators of any accident. 
The probability of a Fuel Handling Accident is unaffected by the 
status of these penetrations.
    The Fuel Handling Accident analyses demonstrate that the maximum 
offsite dose is well [within] the acceptance limits specified in SRP 
[Standard Review Plan] 15.7.4, and the control room dose is within 
the acceptance criteria specified in GDC [General Design Criterion] 
19. Furthermore, the existing analysis results are independent of 
the containment release path, and therefore are unaffected by the 
proposed change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve the addition or 
modification of any plant equipment. Also, the proposed change will 
not alter the design, configuration, or method of operation of the 
plant beyond the standard functional capabilities of the equipment. 
The proposed change involves a Technical Specification change that 
will allow containment penetrations identified under Technical 
Specification 3.9.4(c) to remain open during fuel movement and core 
alterations. Open penetrations are not accident initiators, and will 
not create the possibility of a new kind of accident. Administrative 
controls will be implemented to ensure the capability to close the 
affected containment penetrations in the event of a Fuel Handling 
Accident.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change has the potential to slightly increase the 
post-Fuel Handling Accident dose at the site boundary and in the 
control room. However, the existing analyses take no credit for 
containment of the release, so that the existing analysis results 
will remain bounding.

    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, Mail Stop A-GO-18, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Russell Gibbs.

[[Page 17950]]

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

    Date of amendment request: January 19, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 5.5.9, ``Diesel Fuel Oil Testing 
Program,'' by relocating a reference to a specific American Society for 
Testing and Materials (ASTM) international standard for fuel oil 
testing to licensee-controlled documents, and by adding an alternate 
criteria to the ``clear and bright'' acceptance test for new fuel oil, 
per the consolidated line item improvement process (CLIIP).
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on February 22, 2006 
(71 FR 9179), on possible amendments concerning the CLIIP, including a 
model safety evaluation and a model no significant hazards 
consideration determination. The NRC staff subsequently issued a notice 
of availability of the models for referencing in license amendment 
applications in the Federal Register on April 21, 2006 (71 FR 20735), 
as part of the CLIIP.
    In its application dated January 19, 2007, the licensee affirmed 
the applicability of the following determination.
    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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Requirements to perform testing in 
accordance with applicable ASTM standards are retained in the TS as 
are requirements to perform surveillances of both new and stored 
diesel fuel oil. Future changes to the licensee-controlled document 
will be evaluated pursuant to the requirements of 10 CFR 50.59, 
``Changes, tests and experiments,'' to ensure that such changes do 
not result in more than a minimal increase in the probability or 
consequences of an accident previously evaluated. In addition, the 
``clear and bright'' test used to establish the acceptability of new 
fuel oil for use prior to addition to storage tanks has been 
expanded to recognize more rigorous testing of water and sediment 
content. Relocating the specific ASTM standard references from the 
TS to a licensee-controlled document and allowing a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil will not affect nor degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. In addition, the ``clear and bright'' 
test used to establish the acceptability of new fuel oil for use 
prior to addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. 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 requirements retained in the TS continue to require 
testing of the diesel fuel oil to ensure the proper functioning of 
the DGs.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Instituting the proposed changes will 
continue to ensure the use of applicable ASTM standards to evaluate 
the quality of both new and stored fuel oil designated for use in 
the emergency DGs.
    Changes to the licensee-controlled document are performed in 
accordance with the provisions of 10 CFR 50.59. This approach 
provides an effective level of regulatory control and ensures that 
diesel fuel oil testing is conducted such that there is no 
significant reduction in a margin of safety.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
margin of safety provided by the DGs is unaffected by the proposed 
changes since there continue to be TS requirements to ensure fuel 
oil is of the appropriate quality for emergency DG use. The proposed 
changes provide the flexibility needed to improve fuel oil sampling 
and analysis methodologies while maintaining sufficient controls to 
preserve the current margins of safety.

    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: Russell A. Gibbs

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

    Date of amendment request: December 14, 2006, as supplemented by 
letter dated March 14, 2007.
    Description of amendment request: The proposed amendment would 
modify the technical specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The 
changes are consistent with NRC approved Industry/Technical 
Specification Task Force (TSTF) standard TS change TSTF-372, Revision 
4.
    The proposed amendment includes an administrative change to LCO 
3.0.1 that will clarify that LCO 3.0.7 allows specified TS requirements 
to be suspended during physics tests performed in accordance with TSs 
3.1.8 and 3.1.9. This administrative change will make the CR-3 TSs more 
consistent with the standard TSs and with TSTF-372, Revision 4.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed 
the applicability of the model NSHC determination in its application 
dated April 26, 2006.
    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:


[[Page 17951]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. Entrance into Actions 
or delaying entrance into Actions is not an initiator of any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on the delay time allowed 
before declaring a TS supported system inoperable and taking its 
Conditions and Required Actions are no different than the 
consequences of an accident under the same plant conditions while 
relying on the existing TS supported system Conditions and Required 
Actions. Therefore, the consequences of an accident previously 
evaluated are not significantly increased by this change. Therefore, 
this 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.
    The proposed change allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
does 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 normal plant operations. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
restores an allowance in the pre-ISTS [improved Standard Technical 
Specifications] conversion TS that was unintentionally eliminated by 
the conversion. The pre-ISTS TSs were considered to provide an 
adequate margin of safety for plant operation, as does the post-ISTS 
conversion TS. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    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: Thomas H. Boyce.

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

    Date of amendment request: March 16, 2007.
    Description of amendment request: The proposed amendment would add 
new Technical Specification (TS) requirements for the response times 
associated with a steam generator feedwater pump (SGFP) trip and 
feedwater isolation valve (FIV) closure. The amendment would also 
revise the TS requirements for the containment fan cooler unit (CFCU) 
cooling water flow rate. These changes are associated with a revised 
containment response analysis that credits a SGFP trip and FIV closure 
(on a feedwater regulator valve failure) to reduce the mass/energy 
release to the containment during a main steam line break (MSLB). The 
containment analysis also credits a reduced heat removal capability for 
the CFCUs, allowing a reduction in the required service water (SW) flow 
to the CFCUs.
    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 does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed change establishes response time requirements for 
feedwater isolation and reduced CFCU flow rates to support 
containment analyses to accommodate reduced CFCU heat removal 
capacity. The changes in analysis input assumptions affect plant 
response to an accident and are not accident initiators; therefore, 
they have no bearing on the probability of an accident. The Salem 
FSAR [Final Safety Analysis Report] Chapter 15 accidents which are 
impacted by a change in the CFCU modeling parameters are LOCA [loss-
of-coolant accident] and MSLB mass and energy release Containment 
analyses. The consequences of these postulated accidents are shown 
to be acceptable using assumptions consistent with the proposed 
changes.
    For the LOCA transients, the containment cooling systems are 
considered for three aspects: core response, containment response 
and dose. The core response is most limiting when the containment 
conditions minimize back pressure since this increases the blowdown 
and reduces the effectiveness of the ECCS [emergency core cooling 
system]. The LOCA core response (10 CFR 50.46 [Section 50.46 of 
Title 10 of the Code of Federal Regulations]--PCT [peak cladding 
temperature]) is conservatively biased to minimize the containment 
backpressure such that any safety injection effectiveness is 
minimized (the core becomes the highest resistance flow path). Thus, 
any reduction in the accident capability of the CFCUs has no bearing 
on the LOCA core response.
    The bounding containment integrity analyses are the LBLOCA 
[large-break LOCA] and the MSLB Inside Containment events. The 
containment integrity analysis relies on two heat removal paths to 
maintain containment pressure and temperature conditions. The CFCU 
air-to-water heat exchangers reject containment energy to the SW 
System and the Containment Spray System removes containment energy 
by using spray droplet direct contact heat exchange to transfer the 
energy from the containment ambient to the containment sump, where 
it is transferred out of containment via the RHR [residual heat 
removal] heat exchanger and CCW [component cooling water]/SW 
Systems. Containment integrity analyses for both LOCA and MSLB, 
using input assumptions consistent with the proposed changes, show 
that containment integrity is maintained with reduced CFCU heat 
removal capacity.
    The potential dose impacts due to reduced CFCU heat removal 
capacity are bounded as the design basis assumptions concerning the 
number of operating CFCUs (three of five), and the thermal-hydraulic 
transient operation of the Containment Spray System are unchanged. 
The Salem design basis only credits Containment Spray iodine removal 
effectiveness during the LOCA injection and recirculation phases 
based on a single failure of an entire ESF [engineered safety 
features] train. This assumption results in 3 of 5 CFCUs being 
available to ensure adequate mixing of the containment ambient air 
as well as operation of a single Containment Spray Train, which 
controls containment spray droplet size and pH, as described in 
UFSAR [updated FSAR] Section 6.2.3. As a further conservatism, the 
current LOCA Alternate Source Term (AST) analysis (Calculation S-C-
ZZ-MDC-1945, an interim revision of which was sent to the NRC 
[Nuclear Regulatory Commission] staff for review via letter dated 
September 16, 2004) only credits two CFCUs for mixing. The 
Containment Building and Auxiliary Building leakage rates are 
unaffected by the revised containment analysis as the peak 
containment pressure and temperatures are less than the design basis 
values described in the Salem UFSAR. Therefore, there is no impact 
on offsite dose rates due to the reduced CFCU heat removal capacity.
    One other high energy line break for consideration is the 
rupture of a feedwater line break. From a containment response 
aspect, this event is bounded by the MSLB event, so it is not 
explicitly analyzed (or even discussed in the Salem UFSAR).
    A review of the Salem design basis for AST dose calculations 
shows that the revised Containment Integrity Analysis, WCAP-16503, 
does not challenge any of the assumptions that are part of the AST 
design basis.
    Section 6.2 of the UFSAR indicates that the Appendix J Type A 
containment leak rate test

[[Page 17952]]

pressure is based on the containment design pressure of 47.0 psig, 
not the calculated accident pressure. Since the design pressure 
value bounds the peak pressure calculated in WCAP-16503 and is not 
being changed, the Appendix J testing requirements are not impacted.
    Thus, in conclusion, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The proposed change modifies response 
time requirements for feedwater isolation, and reduces CFCU flow 
rates and heat removal requirements consistent with the new 
containment analysis.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does no