Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 48201-48210 [E5-4403]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 157 (Tuesday, August 16, 2005)]
[Notices]
[Pages 48201-48210]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E5-4403]



[[Page 48201]]

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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 July 22, 2005, to August 4, 2005. The last 
biweekly notice was published on August 2, 2005 (70 FR 44400).

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

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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.

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

    Date of amendments request: June 20, 2005.
    Description of amendments request: The proposed change would revise 
the Technical Specification Surveillance Requirement 3.6.1.6.2 of 
3.6.1.6, ``Suppression Chamber-to-Drywell Vacuum Breakers'' for the 
frequency of functionally testing the suppression chamber-to-drywell 
vacuum breakers.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change revises Surveillance Requirement [SR] 
3.6.1.6.2 to require performance of functional testing of each 
suppression chamber-to-drywell vacuum breaker every 92 days, within 
12 hours after any discharge of steam to the suppression chamber 
from the safety/relief valves, and within 12 hours following an 
operation that causes any of the vacuum breakers to open.
    The proposed change does not involve physical changes to any 
plant structure, system, or component. The suppression chamber-to-
drywell vacuum breakers only provide an accident mitigation 
function. As such, the probability of occurrence for a previously 
analyzed accident is not impacted by the change to the surveillance 
frequency for these components. The consequences of a previously 
analyzed accident are dependent on the initial conditions assumed 
for the analysis, the behavior of the fuel during the analyzed 
accident, the availability of successful functioning of the 
equipment assumed to operate in response to the analyzed event, and 
the setpoints at which these actions are initiated. No physical 
change to suppression chamber-to-drywell vacuum breakers is being 
made as a result of the proposed change, nor does the change alter 
the manner in which the vacuum breakers operate. As a result, no new 
failure modes of the suppression chamber-to-drywell vacuum breakers 
are being introduced. The proposed quarterly surveillance frequency 
for the suppression chamber-to-drywell vacuum breakers is consistent 
with the American Society of Mechanical Engineers (ASME) Code 
frequency for testing these valves, will avoid unnecessary cycling 
and wear of the vacuum breakers, and will improve the reliability of 
the vacuum breakers. Based on this evaluation, there is no 
significant increase in the consequences of a previously analyzed 
event.
    Therefore, the proposed change to the surveillance frequency for 
the suppression chamber-to-drywell vacuum breakers does not involve 
a significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    The proposed change to the surveillance frequency for the 
suppression chamber-to-drywell vacuum breakers does not involve any 
physical alteration of plant systems, structures, or components. No 
new or different equipment is being installed. No installed 
equipment is being operated in a different manner. There is no 
alteration to the parameters within which the plant is normally 
operated or in the setpoints that initiate protective or mitigative 
actions. As a result no new failure modes are being introduced. 
Therefore, the proposed change to the surveillance frequency for the 
suppression chamber-to-drywell vacuum breakers does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed change revises SR 3.6.1.6.2 to require performance 
of functional testing of each vacuum breaker every 92 days, within 
12 hours after any discharge of the steam to the suppression chamber 
from the safety/relief valves, and within 12 hours following an 
operation that causes any of the vacuum breakers to open. The 
operability and functional characteristics of the suppression 
chamber-to-drywell vacuum breakers remains unchanged. The margin of 
safety is established through the design of the plant structures, 
systems, and components, through the parameters within which the 
plant is operated, through the establishment

[[Page 48203]]

of the setpoints for the actuation of equipment relied upon to 
respond to an event, and through the margins contained within the 
safety analyses. The proposed change to the surveillance frequency 
for the suppression chamber-to-drywell vacuum breakers does not 
impact the condition or performance of structures, systems, 
setpoints, and components relied upon for accident mitigation. As 
previously noted, the proposed quarterly surveillance frequency for 
the suppression chamber-to-drywell vacuum breakers is consistent 
with the ASME Code for frequency for testing these vacuum breakers, 
will avoid unnecessary cycling and wear of the vacuum breakers, and 
will improve the reliability of the vacuum breakers. The proposed 
change does not impact any safety analysis assumptions or results. 
Therefore, the proposed change does not result in a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: June 29, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) to revise Surveillance 
Requirements (SR) 3.6.1.3.11 and 3.6.1.3.12 in TS 3.6.1.3, ``Primary 
Containment Isolation Valves (PCIVs).'' Specifically, the proposed 
amendment would revise the combined secondary containment bypass 
leakage rate limit for all bypass leakage paths in SR 3.6.1.3.11 from 
0.05 to 0.10 La and the combined main steam isolation valve 
(MSIV) leakage rate limit for all four main steam lines in SR 
3.6.1.3.12 from 150 to 250 standard cubic feet per hour (scfh).
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The increase in the allowed secondary containment bypass leakage 
limit in SR 3.6.1.3.11 and the increase in the total Main Steam 
Isolation Valve (MSIV) leakage rate limit have been evaluated in a 
revision to the analysis of the Loss of Coolant Accident (LOCA). 
Based on the results of the analysis, it has been demonstrated that, 
with the requested change, the dose consequences of this limiting 
Design Basis Accident (DBA) are within the regulatory guidance 
provided by the NRC [Nuclear Regulatory Commission] for use with the 
AST [alternative source term]. This guidance is presented in 10 CFR 
50.67, Regulatory Guide 1.183, ''Alternative Radiological Source 
Terms For Evaluating Design Basis Accidents At Nuclear Power 
Reactors,'' and Standard Review Plan (SRP) Section 15.0.1. The 
proposed change also updates the design basis value for the Control 
Room Envelope (CRE) unfiltered inleakage based on actual test 
results. This is acceptable because the assumed value in the 
analysis is more than three times the worst case test value. The 
proposed change does not affect the normal design or operation of 
the facility before the accident; rather, it affects leakage limit 
assumptions that constitute inputs to the evaluation of the 
consequences. The radiological consequences of the analyzed LOCA 
have been evaluated using the plant licensing basis for this 
accident. The results conclude that the control room and offsite 
doses remain within applicable regulatory limits. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change in leakage limits does not affect the design, 
functional performance or normal operation of the facility. 
Similarly, it does not affect the design or operation of any 
component in the facility such that new equipment failure modes are 
created. As such the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    This proposed license amendment involves changes in leakage rate 
limits for the secondary containment bypass leakage and MSIV 
leakage. The revised leakage rate limits are used in the LOCA 
radiological analysis in conjunction with the revised CRE unfiltered 
inleakage limit. The analysis has been performed using conservative 
methodologies. Safety margins and analytical conservatisms have been 
evaluated and have been found acceptable. The analyzed LOCA event 
has been carefully selected and margin has been retained to ensure 
that the analysis adequately bounds postulated event scenario. The 
dose consequences of this limiting event are within the acceptance 
criteria presented in 10 CFR 50.67, Regulatory Guide 1.183 and SRP 
Section 15.0.1. The margin of safety is that provided by meeting the 
applicable regulatory limits. The effect of the revision to the 
Technical Specification requirements has been analyzed and doses 
resulting from the pertinent design basis accident have been found 
to remain within the regulatory limits. The change continues to 
ensure that the doses at the exclusion area and low population zone 
boundaries, as well as the control room, are within the 
corresponding regulatory limits. Therefore, the proposed change will 
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 G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: June 8, 2005.
    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 10 CFR 50.65(a)(4). 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 June 8, 2005.
    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

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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 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 a 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 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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: May 31, 2005.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), part 50, section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
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 April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated May 31, 2005.
    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 entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. 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). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, 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 a Margin of Safety.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. 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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.

[[Page 48205]]

    NRC Section Chief: Richard J. Laufer.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: May 24, 2005.
    Description of amendment request: The proposed amendment would 
delete the Technical Specification (TS) temperature limit for the 
safety relief valve (SRV) discharge pipe and the requirements for NRC 
approval of the associated engineering evaluation.
    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. This proposed change deletes an administrative 
requirement for NRC approval of an engineering evaluation to resolve 
a non-conforming and degraded condition that is required by NRC 
Generic Letter 91-18 (GL), Rev. 1, ``Information to Licensees 
Regarding NRC Inspection Manual Section on Resolution of Degraded 
and Nonconforming Conditions''. The SRVs will be maintained 
operable, inspected, and tested to perform their safety function as 
required by the current Specifications and any SRV non-conforming or 
degraded condition will be addressed in accordance with GL 91-18. 
The proposed change also deletes a Note regarding installed two-
stage Target Rock SRVs. The deletion of an administrative 
requirement and the Note does not change the plant response to the 
design basis accident and does not increase the probability of 
inadvertent SRV operation. Therefore, the proposed change does not 
significantly increase the probability or consequences of any 
previously evaluated accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The safety function of the SRVs is to provide 
over-pressure protection of the primary coolant pressure boundary 
and also for the automatic functions to rapidly depressurize the 
primary system to a pressure at which low-pressure cooling systems 
can provide makeup. The proposed change deletes an administrative 
requirement and a Note related to installed two-stage Target Rock 
SRVs, and does not introduce any new modes of equipment operation or 
failure. 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?
    Response: No. The ability of the SRVs to perform their safety 
function is maintained during operation and will continue to be 
tested as required in accordance with TS 3/4.13, Inservice Code 
Testing. The proposed change deletes an administrative requirement 
that is adequately addressed by following GL 91-18, Rev. 1. Deletion 
of an administrative requirement does not reduce the margin of 
safety. 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: May 24, 2005.
    Description of amendment request: The proposed amendment would 
delete the main steam isolation valve (MSIV) twice per week partial 
stroke testing surveillance specified in Technical specification (TS) 
4.7.A.2.b.1.c.
    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. This proposed change deletes the requirement to 
exercise the MSIV's twice per week at power. The MSIVs will continue 
to be full stroke tested by the Inservice Testing Program. The MSIVs 
will continue to be able to perform their accident mitigation 
function. The plant response to the design basis accident will not 
change and the probability of inadvertent MSIV closure will not be 
increased. Therefore, the proposed change does not significantly 
increase the probability or consequences of any previously evaluated 
accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The safety function of the MSIVs is to isolate the 
main steam lines in case of design basis accidents to limit the loss 
of reactor coolant and/or limit the release of radioactive 
materials. The proposed change does not introduce any new modes of 
equipment operation or failure. 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?
    Response: No. The ability of the MSIVs to perform their safety 
function is tested during the MSIV full stroke fast closure test in 
accordance with TS 3.13, Inservice Testing Program. The proposed 
change deletes a high-risk surveillance. Deletion of the high-risk 
surveillance does not reduce the margin of safety. 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: March 7, 2005.
    Description of amendment request: The proposed amendment request 
will add two NRC approved topical report references to the list of 
analytical methods in Technical Specification 5.6.5, ``Core Operating 
Limits Report (COLR),'' that can be used to determine core operating 
limits. The proposed changes are:

    1. Add a NRC previously approved Siemans Power Corporation (SPC) 
topical report reference for determination of fuel assembly critical 
power for previously loaded Global Nuclear Fuel (GNNF) GE14 fuel 
which will be co-resident with reloaded Framatome ANP ATRIUM-10 
fuel.
    2. Add a NRC previously approved Framatome Advanced Nuclear 
Power, Inc. (FRA-ANP) topical report reference for an uprated 
methodology for evaluation of loss coolant accident (LOCA) 
conditions.

    The proposed changes are the result of a redesign to untilize 
Framatome ANP ATRIUM-10 fuel during the Unit 1 Refueling Outage 11 
currently scheduled for February 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:


[[Page 48206]]


    Criterion 1--Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes will add two additional NRC approved 
topical report references to the list of administratively controlled 
analytical methods in Technical Specification (TS) 5.6.5, ``Core 
Operating Limits Report (COLR),'' that can be used to determine core 
operating limits. TS 5.6.5 lists NRC approved analytical methods 
used at LaSalle County Station (LSCS) to determine core operating 
limits.
    LSCS Unit 1 is scheduled to reload Framatome ANP ATRIUM-10 fuel 
during the Unit 1 Refueling Outage 11currently scheduled for 
February 2006. The proposed changes to TS Section 5.6.5 will add 
FRA-ANP methodologies to determine overall core operating limits for 
future core configurations. This change will require the listing of 
additional analytical methods for evaluating LOCA conditions and 
determining the critical power performance of the GE14 fuel. Thus, 
the proposed changes will allow LSCS to use the most recent FRA-ANP 
LOCA methodology for evaluation of ATRIUM-10 fuel and SPC critical 
power correlations to determine the critical power for the GE14 
fuel.
    The addition of approved methods to TS Section 5.6.5 has no 
effect on any accident initiator or precursor previously evaluated 
and does not change the manner in which the core is operated. The 
methods have been reviewed to ensure that the output accurately 
models predicted core behavior, have no effect on the type or amount 
of radiation released, and have no effect on predicted offsite doses 
in the event of an accident. Additionally the methods do not change 
any key core parameters that influence any accident consequences. 
Thus, the proposed changes do not have any effect on the probability 
of an accident previously evaluated.
    The methodology conservatively establishes acceptable core 
operating limits such that the consequences of previously analyzed 
events are not significantly increased.
    The proposed changes in the administratively controlled 
analytical methods do not affect the ability of LSCS to successfully 
respond to previously evaluated accidents and does 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.
    Criterion 2--Does the proposed change create the possibility of 
a new or different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes involve TS 5.6.5 do not affect the 
performance of any LSCS structure, system, or component credited 
with mitigating any accident previously evaluated. The insertion of 
fuel, which has been analyzed with NRC approved methodologies, will 
not affect the control parameters governing unit operation or the 
response of plant equipment to transient conditions. The proposed 
changes do not introduce any new modes of system operation or 
failure mechanism.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Criterion 3--Do the proposed changes involve a significant 
reduction in the margin of safety.
    Response: No.
    The proposed changes will add two additional references to the 
list of administratively controlled analytical methods in TS 5.6.5 
that can be used to determine core operating limits. The proposed 
changes do not modify the safety limits or 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, LSCS has 
determined that the proposed changes provide an equivalent level of 
protection as that currently provided.
    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. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

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

    Date of amendment request: July 5, 2005.
    Description of amendment request: The proposed amendment would 
modify the existing Technical Specification (TS) 3.3.1.3, ``Oscillation 
Power Range Monitor (OPRM) Instrumentation,'' Surveillance Requirement 
(SR) 3.3.1.3.5. Specifically, the thermal power level at which the 
OPRMs are ``not bypassed'' (enabled to perform their design function) 
will be changed from > 28.6 percent rated thermal power to >= 23.8 
percent rated thermal power.
    Plant-specific stability calculations are now required as part of 
the resolution to several generic issues associated with OPRM 
operability. One of the outcomes from this resolution was a change in 
the OPRM enabled region of the power to flow map. The thermal power 
level for enabling the OPRMs for Cycle 10 became > 27.2 percent rated 
thermal power. Since the current TS SR requirement is > 28.6 percent, 
the new TS SR thermal power level value is considered a non-
conservative TS. The Perry Nuclear Power Plant (PNPP) is currently 
requiring the OPRMs to be enabled at >= 23.8 percent thermal power 
level through administrative controls. These controls will remain in 
place until such time that this license amendment is approved 
(reference NRC Administrative Letter 98-10, ``Dispositioning of 
Technical Specifications That Are Insufficient to Assure Plant 
Safety,'' dated December 12, 1998).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change involves the use of a revised thermal power 
level to establish the OPRM enabled region. The OPRM enabled region 
is that area on the power to flow map where the OPRM System is 
activated to detect and suppress potential instability events. If 
reactor operations result in entrance into this region and a core 
instability is detected, the OPRM System will automatically initiate 
a reactor scram. The revised enabled region provides assurance that 
the requirements of 10CFR50, Appendix A, General Design Criteria 10 
and 12 remain satisfied for current and future core designs. Though 
the initiation of instability events are dependent upon thermal 
power levels and core flows, the revision to the enabled region 
thermal power level value does not increase the possibility of such 
an event. Once the OPRMs are enabled, the OPRM System would still 
mitigate an instability event, if detected. The revised enabled 
region does not impact any OPRM detection or mitigation actions for 
instability events.
    The OPRMs are designed to detect and suppress potential 
instability events. As such, the OPRMs are not credited to provide 
any type of detection or mitigation actions for transients or 
accidents described within the PNPP Updated Final Safety Analysis 
Report (USAR) other than instability events. Hence, revising the 
OPRMs enabled region will not impact the transients or accidents 
described within the PNPP Updated Safety Analysis Report (USAR) 
other than instability events.
    Since the OPRMs will be enabled at a thermal power lower than 
analytically required, the potential for additional scrams exists. 
However, since the possibility of an instability event occurring in 
the range between the revised thermal power level and the analytical 
value is remote, the probability of an additional scram from 
occurring is not significantly increased.

[[Page 48207]]

    Therefore, since no significant changes are being made to the 
plant or its design, the probability or the consequences of an 
accident have not increased over those previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change involves the use of a revised thermal power 
level to establish the OPRM enabled region. The use of a revised 
thermal power level to establish the OPRM enabled region does not 
involve a physical modification to any plant system or component, 
including the fuel. The revised enabled region provides assurance 
that the requirements of 10CFR50, Appendix A, General Design 
Criteria 10 and 12 remain satisfied for current and future core 
designs. Though the initiation of instability events are dependent 
upon thermal power levels and core flows, the revision to the 
enabled region thermal power level value does not increase the 
possibility of such an event, or introduce any new or different 
events. Once the OPRMs are enabled, the OPRM System detects and 
mitigates an instability event if detected. The revised enabled 
region does not impact any mitigation actions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change involves the use of a revised thermal power 
level to establish the OPRM enabled region. Once the OPRMs are 
enabled, the OPRM System mitigates an instability event if detected. 
The revised enabled region does not impact any mitigation actions. 
The use of a revised thermal power level to establish the OPRM 
enabled region does not involve a physical modification to any plant 
system or component, including the fuel. The revised enabled region 
provides assurance that the requirements of 10CFR50, Appendix A, 
General Design Criteria 10 and 12 remain satisfied for current and 
future core designs. The revised enabled region restores the margin 
of protection provided by the OPRMs, which had been reduced as fuel 
and core designs have evolved since 1994. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

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

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: May 25, 2005.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), part 50, section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
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 April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated May 25, 2005.
    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 entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. 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). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, 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 a Margin of Safety.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and

[[Page 48208]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Evangelos C. Marinos.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: July 5, 2005.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) would add a reference in TS 5.65.b, 
``Core Operating Limits Report (COLR),'' to permit the use of an 
alternate methodology, VIPRE-D/BWU code/correlation (Virginia Electric 
and Power Company version of the Electric Power Research Institute 
(EPRI) computer code VIPRE [Versatile Internals and Components Program 
for Reactors--EPRI] with the BWU Critical Heat Flux (CHF) 
correlations), to perform thermal-hydraulic analysis to predict CHF and 
Departure from Nucleate Boiling Ratio (DNBR) for the AREVA Advanced 
Mark-BW (AMBW) fuel in the North Anna cores.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The probability of occurrence or the consequences of an 
accident previously evaluated are not significantly increased.
    Neither the code/CHF correlation pair nor the Statistical DNBR 
Evaluation Methodology make any contribution to the potential 
accident initiators and thus cannot increase the probability of any 
accident. Further, since both the deterministic and statistical DNBR 
limits meet the required design basis of avoiding DNB with 95% 
probability at a 95% confidence level, the use of the new code/
correlation and Statistical DNBR Evaluation Methodology do not 
increase the potential consequences of any accident. Finally the 
addition of a full core DNB design limit provides increased 
assurance that the consequences of a postulated accident which 
included radioactive release would be minimized because the overall 
number of rods in DNB would not exceed the 0.1% level. All the 
pertinent evaluations to be performed as part of the cycle specific 
reload safety analysis to confirm that the existing safety analyses 
remain applicable have been performed with VIPRE-D/BWU and found to 
be acceptable. The use of a different code/correlation pair will not 
increase the probability of an accident because plant systems will 
not be operated in a different manner, and system interfaces will 
not change. The use of the VIPRE-D/BWU code/correlation pair will 
not result in a measurable impact on normal operating plant 
releases, and will not increase the predicted radiological 
consequences of accidents postulated in the UFSAR [Updated Final 
Safety Analysis Report]. Therefore, neither the probability of 
occurrence nor the consequences of any accident previously evaluated 
is significantly increased.
    2. The possibility for a new or different type of accident from 
any accident previously evaluated is not created.
    The use of VIPRE-D/BWU and its applicable fuel design limits for 
DNBR does not impact any of the applicable design criteria and all 
pertinent licensing basis criteria will continue to be met. 
Demonstrated adherence to these standards and criteria precludes new 
challenges to components and systems that could introduce a new type 
of accident. Setpoint safety analysis evaluations have demonstrated 
that the use of VIPRE-D/BWU is acceptable. All design and 
performance criteria will continue to be met and no new single 
failure mechanisms will be created. The use of VIPRE-D/BWU code/
correlation or the Statistical DNBR Evaluation Methodology does not 
involve any alteration to plant equipment or procedures that would 
introduce any new or unique operational modes or accident 
precursors. Therefore, the possibility for a new or different kind 
of accident from any accident previously evaluated is not created.
    3. The margin of safety is not significantly reduced. North Anna 
Technical Specification 2.1 specifies that any DNBR limit 
Established by any used code/correlation must provide at least 95% 
non-DNB probability at a 95% confidence level. The use of VIPRE-D/
BWU with the SDLs [Statistical Design Limits] listed in this package 
provides that protection, just as LYNXT/BWU [LYNXT thermal-hydraulic 
computer code with the AREVA BWU CHF correlations] and applicable 
SDLs did. The required DNBR margin of safety for the North Anna 
Nuclear units, which in this case is the margin between the 95/95 
DNBR limit and clad failure, is therefore not reduced. Therefore, 
the margin of safety as defined in the Bases to the North Anna Units 
1 and 2 Technical Specifications is not significantly reduced.

    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: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: Evangelos C. Marinos.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: July 14, 2005.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) would correct two errors in the units of 
measure used to determine the Overtemperature [Delta]T Function 
Allowable Value.
    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 changes involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the UFSAR [Updated Final Safety Analysis Report]. The proposed 
changes correct errors in the unit designations used in the 
f1([Delta]I) equation. The actual numerical values of 
f1([Delta]I) calculated by the equation remain the same, 
only the units applied to the value are changed. The Overtemperature 
[Delta]T function allowable values are utilized by the Reactor Trip 
System (RTS) instrumentation to prevent reactor operation in 
conditions outside the range considered for accident analyses. The 
proposed changes will not alter the allowable values used by the RTS 
instrumentation. The Overtemperature [Delta]T allowable value is not 
an initiator to any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not 
significantly increased. As the Overtemperature [Delta]T allowable 
value is not changed, the probability or consequences of an accident 
previously evaluated is not significantly increased.
    2. Do changes create the possibility of a new or different kind 
of accident from a