Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 26995-27010 [06-4243]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 89 (Tuesday, May 9, 2006)]
[Notices]
[Pages 26995-27010]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-4243]


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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

[[Page 26996]]

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 April 14, 2006 to April 27, 2006. The last 
biweekly notice was published on April 25, 2006 (71 FR 23952).

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 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

[[Page 26997]]

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.

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

    Date of amendment request: January 26, 2006.
    Description of amendment request: The proposed amendment would 
update the list of Nuclear Regulatory Commission-approved documents 
specified in the Technical Specifications that describe the analytical 
methods used to determine the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment adds a new document (No. 16) to TS 
6.9.1.8 b to complement the list of documents used to determine the 
core operating limits. These documents have been previously reviewed 
and approved by the NRC. It also changes the word ``minimum'' to 
``maximum'' in TS 5.3.1 to correctly state the limit on nominal 
average enrichment of reload fuel. This change restores TS 5.3.1 
wording to the wording previously approved by the NRC in Amendment 
274. The proposed changes do not modify any plant equipment and do 
not impact any failure modes that could lead to an accident. 
Additionally, the proposed changes have no effect on the consequence 
of any analyzed accident since the changes do not affect the 
function of any equipment credited for accident mitigation. Based on 
this discussion, the proposed amendment does not increase 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 changes do not modify any plant equipment and there 
is no impact on the capability of existing equipment to perform its 
intended functions. No system setpoints are being modified and no 
changes are being made to the method in which plant operations are 
conducted. No new failure modes are introduced by the proposed 
change. The proposed amendment does not introduce accident 
initiators or malfunctions that would cause a new or different kind 
of accident. 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: No.
    The proposed amendment adds a new document (No. 16) to TS 
6.9.1.8 b to complement the list of documents used to determine the 
core operating limits. These documents have been previously reviewed 
and approved by the NRC. It also changes the word ``minimum'' to 
``maximum'' in TS 5.3.1 to correctly state the limit on nominal 
average enrichment of reload fuel. This change restores TS 5.3.1 
wording to the wording previously approved by the NRC in Amendment 
274. The proposed changes have no impact on plant equipment 
operation. The proposed changes do not revise any setpoints nor do 
they change the acceptance criteria used 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 Branch Chief: Darrell J. Roberts.

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

    Date of amendment request: March 28, 2006.
    Description of amendment request: The proposed amendment would 
delete the license condition, Section 2.F of Facility Operating License 
No. NPF-49, which requires reporting of violations of the requirements 
in Section 2.C of Facility Operating License No. NPF-49. The change is 
consistent with the notice published in the Federal Register on 
November 4, 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:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.

[[Page 26998]]

    The proposed change involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase 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 is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the change 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: No.
    The proposed change deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.

    Based on the above, the NRC staff proposes that the change presents 
no significant hazards consideration under the standards set forth in 
10 CFR 50.92(c).
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Branch Chief: Darrell J. Roberts.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: June 15, 2005.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to eliminate the out of date 
requirements associated with the completion of the Keowee Refurbishment 
modifications on both Keowee Hydro Units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    The proposed change to the Oconee Technical Specification (TS) 
3.8.1 removes out of date requirements associated with temporary 
extensions to Required Action (RA) Completion Times (CTs) that are 
no longer applicable because of the completion of the Keowee 
Refurbishment modifications on both KHUs. The proposed change also 
removes a Facility Operating License (FOL) License Condition that is 
no longer needed since the associated TS change is no longer 
applicable. As such, the proposed change is administrative. No 
actual plant equipment, operating practices, or accident analyses 
are affected by this change. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated:
    The proposed change to the Oconee TSs and FOLs removes 
requirements associated with a temporary extension of TS 3.8.1 RA 
CTs that are no longer applicable because of the completion of the 
Keowee Refurbishment modifications on both KHUs. As such, the 
proposed changes are administrative. No actual plant equipment, 
operating practices, or accident analyses are affected by this 
change. No new accident causal mechanisms are created as a result of 
this change. The proposed change does not impact any plant systems 
that are accident initiators; neither does it adversely impact any 
accident mitigating systems. Therefore, this change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not adversely affect any plant safety 
limits, set points, or design parameters. The change also does not 
adversely affect the fuel, fuel cladding, Reactor Coolant System, or 
containment integrity. The proposed change eliminates requirements 
that are no longer applicable and is administrative in nature. 
Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: April 17, 2006.
    Description of amendment request: The proposed change allows a 
delay time for entering a supported system technical specification (TS) 
when the inoperability is due solely to an inoperable snubber, if risk 
is assessed and managed consistent with the program in place for 
complying with the requirements of paragraph 50.65(a)(4) of Title 10 of 
the Code of Federal Regulations (10 CFR). Limiting Condition for 
Operation (LCO) 3.0.8 is added to the TS to provide this allowance and 
define the requirements and limitations for its use.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
November 24, 2004 (69 FR 68412), on possible amendments concerning 
TSTF-372, including a model safety evaluation and model no significant 
hazards consideration (NSHC) determination, using the consolidated line 
item improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on May 4, 2005 (70 FR 23252). The 
licensee affirmed the applicability of the following NSHC determination 
in its application dated April 17, 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:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast 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 
allowance provided by proposed LCO 3.0.8 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 3.0.8. 
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.

Criterion 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 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 inoperable snubbers, if risk is 
assessed and

[[Page 26999]]

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.

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

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG [Regulatory Guide] 1.177. A bounding risk 
assessment was performed to justify the proposed TS changes. [The 
proposed LCO 3.0.8 defines limitations on the use of the provision 
and includes a requirement for the licensee to assess and manage the 
risk associated with operation with an inoperable snubber.] The net 
change to the margin of safety is insignificant. 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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: March 20, 2006.
    Description of amendment request: The proposed change removes 
Arkansas Nuclear One, Unit 2 reactor coolant system (RCS) structural 
integrity requirements contained in Technical Specification (TS) 
3.4.10.1. The proposed change is consistent with NUREG-1432, ``Standard 
Technical Specifications--Combustion Engineering Plants,'' Revision 
3.1.
    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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to remove the RCS structural integrity 
controls from the TSs does not impact any mitigation equipment or 
the ability of the RCS pressure boundary to fulfill any required 
safety function. Since no accident mitigation or initiators are 
impacted by this change, no design basis accidents are affected.
    Therefore, the proposed change does 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 previously evaluated?
    Response: No.
    The proposed change will not alter the plant configuration or 
change the manner in which the plant is operated. No new failure 
modes are being introduced by the proposed change.
    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 
the margin of safety?
    Response: No.
    Removal of TS 3.4.10.1 from the TSs does not reduce the controls 
that are required to maintain the RCS pressure boundary for ASME 
Code [American Society of Mechanical Engineers' Boiler and Pressure 
Vessel Code] Class 1, 2, or 3 components. No equipment or RCS safety 
margins are impacted due to the proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: January 27, 2006.
    Description of amendment request: The proposed amendment involves 
changes to Technical Specifications Section 3/4 9.1, ``Boron 
Concentration,'' Section 3/4 9.14, ``Spent Fuel Storage,'' and Section 
3/4 5.5.1, ``Fuel Storage Criticality.'' The proposed license amendment 
removes reliance on Boraflex as a neutron absorber in Turkey Point 
Units 3 and 4 spent fuel pool storage racks. To preclude continued loss 
of reactivity margin due to the ongoing degradation of Boraflex, the 
neutron absorbing function currently performed by Boraflex will be 
replaced by some combination of rod cluster control assemblies, Metamic 
rack inserts, and administrative controls that require mixing higher 
reactivity fuel with lower-reactivity fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    No. Operation in accordance with proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The proposed amendments do not 
change or modify the fuel, fuel handling processes, spent fuel 
storage racks, number of fuel assemblies that may be stored in the 
spent fuel pool (SFP), decay heat generation rate, or the spent fuel 
pool cooling and cleanup system. The proposed amendment was 
evaluated for impact on the following previously evaluated events 
and accidents:

a. A fuel handling accident (FHA),
b. A cask drop accident,
c. A fuel mispositioning event,
d. A spent fuel pool boron dilution event,
e. A seismic event, and
f. A loss of spent fuel pool cooling event.

    The probability of a FHA is not significantly increased because 
implementation of the proposed amendment will employ the same 
equipment and process to handle fuel assemblies that is currently 
used. Also, tests have confirmed that the Metamic inserts can be 
installed and removed without damaging the host fuel assemblies. The 
FHA radiological consequences are not increased because the 
radiological source term of a single fuel assembly will remain 
unchanged. Therefore, the proposed amendments do not significantly 
increase the probability or consequences of a FHA.
    The proposed amendments do not increase the probability of 
dropping a fuel transfer cask because they do not introduce any new 
heavy loads to the SFP and do not affect heavy load handling 
processes. Also, the insertion of Metamic rack inserts does not 
increase the consequences of the cask drop accident because the 
radiological source term of that accident is developed from a non-
mechanistically derived quantity of damaged fuel stored in the spent 
fuel pool. Therefore, the proposed amendments do not significantly 
increase the probability or consequences of a cask drop accident.
    Operation in accordance with the proposed amendment will not 
change the probability of a fuel mispositioning event because fuel 
movement will continue to be controlled by approved fuel handling 
procedures. These procedures continue to require identification

[[Page 27000]]

of the initial and target locations for each fuel assembly that is 
moved. The consequences of a fuel mispositioning event are not 
changed because the reactivity analysis demonstrates that the same 
subcriticality criteria and requirements continue to be met for the 
worst-case fuel mispositioning event.
    Operation in accordance with the proposed amendment will not 
change the probability of a boron dilution event because the systems 
and events that could affect spent fuel soluble boron are unchanged. 
The consequences of a boron dilution event are unchanged because the 
proposed amendment reduces the soluble boron requirement below the 
currently required value and the maximum possible water volume 
displaced by the inserts is an insignificant fraction of the total 
spent fuel pool water volume.
    Operation in accordance with the proposed amendment will not 
change the probability of a seismic event, which is an Act of God. 
The consequences of a seismic event are not significantly increased 
because the forcing functions for seismic excitation are not 
increased and because the mass of storage racks with Metamic inserts 
is not appreciably increased. Seismic analyses demonstrate adequate 
stress levels in the storage racks when inserts are installed.
    Operation in accordance with the proposed amendment will not 
change the probability of a loss of SFP cooling event because the 
systems and events that could affect SFP cooling are unchanged. The 
consequences are not significantly increased because there are no 
changes in the SFP heat load or SFP cooling systems, structures or 
components. Furthermore, conservative analyses indicate that the 
current design requirements and criteria continue to be met with the 
Metamic inserts installed.
    Based on the above, it is concluded that the proposed amendments 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Would operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    No. Operation in accordance with the proposed amendments do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. The proposed amendments do not 
change or modify the fuel, fuel handling processes, spent fuel 
racks, number of fuel assemblies that may be stored in the pool, 
decay heat generation rate, or the spent fuel pool cooling and 
cleanup system. The effects of operating with the proposed amendment 
are listed below. The proposed amendments were evaluated for the 
potential of each effect to create the possibility of a new or 
different kind of accident:

a. Addition of inserts to the spent fuel storage racks,
b. New storage patterns,
c. Additional weight from the inserts,
d. Insert movement above spent fuel, and
e. Displacement of fuel pool water by the inserts.

    Each insert will be placed between a fuel assembly and the 
storage cell wall, taking up some of the space available on two 
sides of the fuel assembly. Tests confirm that the insert can be 
installed and removed without damaging the fuel assembly. Analyses 
demonstrate that the presence of the inserts does not adversely 
affect spent fuel cooling, seismic capability, or subcriticality. 
The aluminum (alloy 6061) and boron carbide materials of 
construction have been shown to be compatible with nuclear fuel, 
storage racks and spent fuel pool environments, and generate no 
adverse material interactions. Therefore, placing the inserts into 
the spent fuelpool storage racks can not cause a new or different 
kind of accident.
    Operation with the proposed fuel storage patterns will not 
create a new or different kind of accident because fuel movement 
will continue to be controlled by approved fuel handling procedures. 
These procedures continue to require identification of the initial 
and target locations for each fuel assembly that is moved. There are 
no changes in the criteria or design requirements pertaining to 
spent fuel safety, including subcriticality requirements, and 
analyses demonstrate that the proposed storage patterns meet these 
requirements and criteria with adequate margins. Therefore, the 
proposed storage patterns can not cause a new or different kind of 
accident.
    Operation with the added weight of the Metamic inserts will not 
create a new or different accident. The net effect of the adding the 
maximum number of inserts is to add less than one percent to the 
weight of the loaded racks. Furthermore, the analyses of the racks 
with Metamic inserts installed demonstrate that the stress levels in 
the rack modules continue to be considerably less than allowable 
stress limits. Therefore, the added weight from the inserts can not 
cause a new or different kind of accident.
    Operation with the insert allowed to move above spent fuel will 
not create a new or different kind of accident. The insert with its 
handling tool weighs considerably less than the weight of a single 
fuel assembly. Single fuel assemblies are routinely moved safely 
over spent fuel assemblies and the same level of safety in design 
and operation will be maintained when moving the inserts. 
Furthermore, the effect of a dropped insert to block the top of a 
storage cell has been evaluated in thermal-hydraulic analyses. 
Therefore, the movement of inserts can not cause a new or different 
kind of accident.
    Whereas the installed rack inserts will displace a very small 
fraction of the fuel pool water volume and impose a very small 
reduction in operator response time to previously-evaluated SFP 
accidents, the reduction will not promote a new or different kind of 
accident. Also, displacement of water along two sides of a stored 
fuel assembly may have some local reduction in the peripheral 
cooling flow; however, this effect would be small compared to the 
flow induced through the fuel assembly and would in no way promote a 
new or different kind of accident.
    Based on the above, it is concluded that operation with the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Would operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    No. Operation of the facility in accordance with the proposed 
amendment does not significantly reduce the margin of safety. The 
proposed change was evaluated for its effect on current margins of 
safety related to criticality, structural integrity, and spent fuel 
heat removal capability. The margin of safety for subcriticality 
required by 10 CFR 50.68(b)(4) is unchanged. New criticality 
analysis confirms that operation in accordance with the proposed 
amendment continues to meet the required subcriticality margins. 
Also, the margin of safety for SFP soluble boron concentration is 
actually increased because new analyses require less soluble boron 
than is currently required, and much less than the value required by 
Technical Specifications. The structural evaluations for the racks 
and spent fuel pool with Metamic inserts installed show that the 
rack and spent fuel pool are unimpaired by loading combinations 
during seismic motion, and there is no adverse seismic-induced 
interaction between the rack and Metamic inserts.
    The proposed change does not affect spent fuel heat generation 
or the spent fuel cooling systems. A conservative analysis indicates 
that the design basis requirements and criteria for spent fuel 
cooling continue to be met with the Metamic inserts in place, and 
displacing coolant. Thermal hydraulic analysis of the local effects 
of an installed rack insert blocking peripheral flow show a small 
increase in local water and fuel clad temperatures, but will remain 
within acceptable limits including no departure from nucleate 
boiling.
    Based on these evaluations, operating the facility with the 
proposed amendment 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Michael L. Marshall, Jr.

Nuclear Management Company, LLC, Docket No. 50-306, Prairie Island 
Nuclear Generating Plant, Unit 2, Goodhue County, Minnesota

    Date of amendment request: March 13, 2006.
    Description of amendment request: The proposed amendment would 
involve revision of the surveillance test load in Technical 
Specification (TS) 3.8.1, ``AC Sources--Operating,'' Surveillance 
Requirement (SR) 3.8.1.3. This license amendment request proposes to 
revise SR 3.8.1.3 to require

[[Page 27001]]

testing D5 and D6 monthly at or above 4000 kW to demonstrate TS 
operability. In addition to the TS required testing, NMC will continue 
monthly operation at or above 90 percent of the emergency diesel 
generator (EDG) rated load to assist in early identification of 
degraded EDG capabilities which could prevent performance of their 
safety function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes to reduce the Prairie 
Island Nuclear Generating Plant Unit 2 emergency diesel generator's 
monthly test loading which demonstrates Technical Specification 
operability. The proposed test load will continue to assure that 
both Unit 2 emergency diesel generators have the capacity and the 
capability to assume the maximum auto-connected loads for Unit 2.
    The emergency diesel generators are required to be operable in 
the event of a design basis accident coincident with a loss of 
offsite power to mitigate the consequences of the accident. They are 
also the alternate AC source for a station blackout on the other 
Prairie Island Nuclear Generating Plant unit. The emergency diesel 
generators are not accident initiators and therefore this change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The accident analyses assume that at least one safeguards bus is 
provided with power either from the offsite sources or the emergency 
diesel generators. The Technical Specification changes proposed in 
this license amendment request will continue to assure that both 
Unit 2 emergency diesel generators have the capacity and the 
capability to assume the maximum auto-connected loads for Unit 2. 
Thus, the changes proposed in this license amendment request do not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    The changes proposed in this license amendment do not involve a 
significant increase 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 accident previously evaluated?
    Response: No.
    This license amendment request proposes to reduce the Prairie 
Island Nuclear Generating Plant Unit 2 emergency diesel generator's 
monthly test loading which demonstrates Technical Specification 
operability. The proposed test load will continue to assure that 
both Unit 2 emergency diesel generators have the capacity and the 
capability to assume the maximum auto-connected loads for Unit 2.
    The proposed Technical Specification changes do not involve a 
change in the plant design, system operation, or the use of the 
emergency diesel generators. The proposed changes allow the 
emergency diesel generator to be tested at a reduced load which 
envelopes the required safety function loads and continues to 
demonstrate the capability and capacity of the emergency diesel 
generators to perform their required functions. There are no new 
failure modes or mechanisms created due to testing the emergency 
diesel generators at the proposed test loading. Testing of the 
emergency diesel generators at the proposed test loading does not 
involve any modification in the operational limits or physical 
design of plant systems. There are no new accident precursors 
generated due to the proposed test loading.
    The Technical Specification changes proposed in this license 
amendment do not create the possibility of a new or different kind 
of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment request proposes to reduce the Prairie 
Island Nuclear Generating Plant Unit 2 emergency diesel generator's 
monthly test loading which demonstrates Technical Specification 
operability. The proposed test load will continue to assure that 
both Unit 2 emergency diesel generators have the capacity and the 
capability to assume the maximum auto-connected loads for Unit 2.
    The proposed Technical Specification changes will continue to 
demonstrate that the emergency diesel generators meet the Technical 
Specification definition of operability, that is, the proposed 
testing will demonstrate that the emergency diesel generators will 
perform their safety function and the necessary emergency diesel 
generator attendant instrumentation, controls, cooling, lubrication 
and other auxiliary equipment required for the emergency diesel 
generators to perform their safety function loads are also tested at 
this loading. The proposed testing will also continue to demonstrate 
the capability and capacity of the emergency diesel generators to 
supply the required Unit 2 loss of offsite power coincident with 
Unit 1 station blackout loads. Since the proposed surveillance 
testing will continue to demonstrate operability, and the capability 
and capacity to supply their required Unit 2 loss of offsite power 
coincident with Unit 1 station blackout loads, the proposed 
Technical Specification changes do not involve a significant 
reduction in a margin of safety.
    The Technical Specification changes proposed in this license 
amendment do not involve a significant reduction in a margin of 
safety.
    Based on the above, the Nuclear Management Company concludes 
that the proposed amendment 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.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: February 1, 2006.
    Description of amendment request: The proposed amendment would 
clarify the Technical Specification (TS) testing frequency for the 
Surveillance Requirements (SRs) in TS 3.1.4, ``Control Rod Scram 
Times.''
    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 control rod hydraulic scram insertion system is not an 
initiator to any accident sequence analyzed in the Final Safety 
Analysis Report (FSAR). The changes do not involve any physical 
change to structures, systems, or components (SSCs) and do not alter 
the method of operation or control of SSCs. The current assumptions 
in the safety analysis regarding accident initiators and mitigation 
of accidents (including assumed scram insertion times) are 
unaffected by these changes. No additional failure modes or 
mechanisms are being introduced and the likelihood of previously 
analyzed failures remains unchanged.
    Operation in accordance with the proposed Technical 
Specification (TS) ensures that the control rods and associated 
scram insertion function remain capable of performing the function 
as described in the FSAR [Final Safety Analysis Report]. Therefore, 
the mitigative scram functions will continue to provide the 
protection assumed by the analysis.
    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

[[Page 27002]]

accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints affected by this change at which protective or 
mitigative actions are initiated. This change will not alter the 
manner in which equipment operation is initiated, nor will the 
functional demands on credited equipment be changed. No alterations 
in the procedures that ensure the plant remains within analyzed 
limits are being proposed, and no changes are being made to the 
procedures relied upon to respond to an off-normal event as 
described in the FSAR. As such, no new failure modes are being 
introduced. The change does not alter assumptions made in the safety 
analysis and licensing basis.
    [Therefore, 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 margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. Operation in accordance with the proposed TS ensures 
that the control rod scram insertion system remains capable of 
performing the function as described in the FSAR. Sufficiently rapid 
insertion of control rods following certain accidents (scram time) 
will prevent fuel damage, and thereby maintain a margin of safety to 
fuel damage. No change is being made to the required insertion rate 
specified in plant Technical Specifications. Clarifying when control 
rod insertion times must be verified following movement of fuel 
assemblies, without actually changing the requirement (verification 
of insertion times will continue to be required whenever work that 
might impact the rod insertion time is done), does not reduce the 
margin of safety related to fuel damage.
    Therefore, the change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: October 7, 2005.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to clarify certain 
requirements during fuel movement and core alterations. The amendment 
would make the TSs consistent with the NRC-approved Revision 2 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-51, ``Revise Containment 
Requirements During Handling Irradiated Fuel and Core Alterations,'' 
and NUREG-1433, ``Standard Technical Specifications General Electric 
Plants, BWR [boiling water reactor]/4.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously analyzed?
    Response: No.
    The proposed changes would revise Technical Specifications (TS) 
3.6.5.3.1, FRVS [filtration, recirculation and ventilation system] 
Ventilation System, and 3.6.5.3.2, FRVS Recirculation System, ACTION 
b from, ``* * * containment or operations * * * '' to read ``* * * 
containment and operations * * * '' to be consistent with NUREG-
1433, ``Standard Technical Specifications General Electric Plants, 
BWR/4'' (STS). Technical Specification 3.7.1.2, Service Water, and 
3.8.3.2, Distribution--Shutdown, require the addition of 
``recently'' to modify irradiated fuel consistent with NRC-approved 
Revision 2 to Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-51, ``Revise 
Containment Requirements During Handling Irradiated Fuel and Core 
Alterations.'' Technical Specifications 3.8.1.2, A.C. Sources--
Shutdown, 3.8.2.2, DC Sources--Shutdown, and 3.8.3.2, Distribution--
Shutdown, require that ``CORE ALTERATIONS'' be added to ACTION a.
    The proposed changes associated with the fuel handling accident 
(FHA) do not involve a change to structures, components, or systems 
that would affect the probability of an accident previously 
evaluated in the Hope Creek Updated Final Safety Analysis Report 
(UFSAR). The FHA for Hope Creek is defined as a drop of a fuel 
assembly over irradiated assemblies in the reactor core 24 hours 
after reactor shutdown. 10 CFR 50.67, ``Accident Source Term'' 
(AST), was used to evaluate the dose consequences of a postulated 
accident. The FHA has been analyzed without credit for Secondary 
Containment; Filtration, Recirculation and Ventilation System 
(FRVS); and CREF [control room emergency filtration] system. The 
resultant radiological consequences are within the acceptance 
criteria set forth in 10 CFR 50.67 and Regulatory Guide (RG) 1.183. 
This amendment does not alter the methodology or equipment used in 
fuel handling operations. The equipment hatch, personnel air locks, 
other containment penetrations, or any component thereof is not an 
accident initiator. Actual fuel handling operations are not affected 
by the proposed changes.
    Consequently the probability of a previously analyzed FHA is not 
affected by the proposed amendment. No other accident initiator is 
affected by the proposed changes.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
radiological 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 analyzed?
    Response: No.
    The proposed changes would revise TS 3.6.5.3.1, FRVS Ventilation 
System and 3.6.5.3.2, FRVS Recirculation System, ACTION b from, ``* 
* * containment or operations * * * '' to read ``* * * containment 
and operations * * * '' to be consistent with NUREG-1433, Standard 
Technical Specifications General Electric Plants, BWR/4'' (STS). TS 
3.7.1.2, Service Water, and 3.8.3.2, Distribution--Shutdown, require 
the addition of ``recently'' to modify irradiated fuel consistent 
with NRC-approved Revision 2 to Technical Specification Task Force 
(TSTF) Standard Technical Specification Change Traveler, TSTF-51, 
``Revise Containment Requirements During Handling Irradiated Fuel 
and Core Alterations.'' TS 3.8.1.2 A.C. Sources--Shutdown, 3.8.2.2, 
D.C. Sources--Shutdown, and 3.8.3.2, Distribution--Shutdown, require 
that ``CORE ALTERATIONS'' be added to ACTION a.
    The proposed amendment will not create the possibility of a new 
or different type of accident from any accident previously evaluated 
because changes to the allowable activity in the primary and 
secondary systems do not result in changes to the design or 
operation of these systems. The evaluation of the proposed changes 
indicates that all design standard and applicable safety criteria 
limits are met. Equipment important to safety will continue to 
operate as designed. Component integrity is not challenged. The 
changes do not result in any event previously deemed incredible 
being made credible. The changes do not result in more adverse 
conditions or result in any increase in the challenges to safety 
systems. The systems affected by the changes are used to mitigate 
the consequences of a potential accident and would 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 the margin 
of safety?
    Response: No.
    The proposed changes would revise TS 3.6.5.3.1, FRVS Ventilation 
System and 3.6.5.3.2 FRVS Recirculation System, ACTION b from ``* * 
* containment or operations * * * '' to read ``* * * containment and 
operations * * * '' to be consistent with NUREG-1433, ``Standard 
Technical Specifications General Electric Plants, BWR/4'' (STS). TS 
3.7.1.2, Service Water, and 3.8.3.2, Distribution--Shutdown, require 
the addition of ``recently'' to modify irradiated fuel consistent 
with NRC approved Revision 2 to Technical Specification Task

[[Page 27003]]

Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
51, ``Revise Containment Requirements During Handling Irradiated 
Fuel and Core Alterations.'' TS 3.8.1.2 A.C. Sources--Shutdown, 
3.8.2.2 D.C. Sources--Shutdown, and 3.8.3.2 Distribution--Shutdown, 
require that ``CORE ALTERATIONS'' be added to ACTION a.
    The proposed changes revise the TS operational conditions where 
specific activities represent situations during which significant 
radioactive releases can be postulated. These operational conditions 
are consistent with the design basis analysis and are established 
such that the radiological consequences remain at or below the 
regulatory guidelines. Safety margins and analytical conservatisms 
are retained to ensure that the analysis adequately bounds all 
postulated event scenarios. The proposed TS continue to ensure that 
the total effective dose equivalent (TEDE) for the control room 
(CR), the exclusion area boundary (EAB), and low population zone 
(LPZ) boundaries are below the corresponding acceptance criteria 
specified in 10 CFR 50.67 and RG 1.183.
    Therefore, these 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Darrell J. Roberts.

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

    Date of amendment request: February 23, 2006.
    Description of amendment request: The amendment would revise the 
Operating License Condition 2.C.(6), ``Fuel Storage and Handling,'' to 
clarify that the condition does not apply to Nuclear Regulator 
Commission (NRC)-approved dry spent fuel storage systems. The current 
condition states no more than a total of three fuel assemblies shall be 
out of approved shipping containers, fuel assembly storage racks or the 
reactor at any one time.
    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 is a clarification to the Hope Creek 
operating license to recognize that the dry spent fuel storage 
system used at the ISFSI [independent spent fuel storage 
installation] is licensed separately by the NRC under 10 CFR part 
72. The change does not affect any SSCs [structure, systems and 
components] used to operate the reactor or produce electrical power. 
The change also does not affect SSCs used to shut down the reactor, 
maintain it in a safe shutdown condition, or mitigate accidents.
    The dry storage cask system design is supported by an NRC-
approved criticality analysis that demonstrates the system will 
remain safely subcritical under all normal, off-normal, and credible 
accident conditions applicable to the dry spent fuel storage system, 
as defined in the cask CoC holder's 10 CFR part 72 licensing basis. 
Dry spent fuel storage system loading operations are not addressed 
in any Part 50 accident as described in Chapter 15 of the HCGS [Hope 
Creek Generating Station] FSAR [final safety analysis report]. Dry 
spent fuel storage system loading in the spent fuel pool is governed 
by procedures that are consistent with the requirements in the HI-
STORM 100 System 10 CFR part 72 FSAR. Heavy load handling inside the 
Part 50 facility associated with cask loading is conducted in 
accordance with procedures that comply with the site's existing 
heavy load control program. Because this change does not affect 
PSEG's [PSEG Nuclear, LLC] heavy load handling procedures and all 
structures, systems and components used for cask handling will meet 
the existing commitments to NUREG-0612, a cask drop event remains 
non-credible as currently described in HCGS FSAR Section 15.7.5.
    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.
    The proposed change is a clarification to the Hope Creek 
operating license to recognize that the dry spent fuel storage 
system is licensed separately by the NRC under 10 CFR part 72. The 
change does not affect any SSCs used to operate the reactor or 
produce electrical power. The change also does not affect SSCs used 
to shut down the reactor, maintain it in a safe shutdown condition, 
or mitigate accidents.
    The dry spent fuel storage system design is supported by an NRC-
approved criticality analysis that demonstrates the system will 
remain safely subcritical under all normal, off-normal, and credible 
accident conditions, as defined in the cask CoC holder's 10 CFR part 
72 licensing basis. Dry spent fuel storage system loading in the 
spent fuel pool is governed by procedures that are consistent with 
the requirements in the HI-STORM 100 System 10 CFR 72 FSAR. Heavy 
load handling inside the Part 50 facility associated with cask 
loading is conducted in accordance with procedures that comply with 
the site's existing heavy load control program. Because this change 
does not affect PSEG's heavy load handling procedures and all 
structures, systems and components used for cask handling will meet 
the existing commitments to NUREG-0612, a cask drop event remains 
non-credible as currently described in HCGS FSAR Section 15.7.5.
    Therefore, the proposed change will 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 
the margin of safety?
    Response: No.
    The proposed change is a cl