Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72667 

1 p.m.—Meeting with the Advisory 

Committee on Reactor Safeguards 

(ACRS), (Contact: John Larkins, 

301–415–7360). 

This meeting will be webcast live at 

the Web address: http://www.nrc.gov. 

Week of December 12, 2005—Tentative 

Monday, December 12, 2005. 

8:50 a.m.—Affirmation Session 

(Public Meeting) (Tentative), a. 

Exelon Generation Company, LLC 

(Early Site Permit for Clinton Site). 

(Tentative). 

9 a.m.—Discussion of Security Issues 

(Closed—Ex. 1). 

Wednesday, December 14, 2005. 

1:30 p.m.—Discussion of Security 

Issues (Closed—Ex. 1). 

Thursday, December 15, 2005. 

1:30 p.m.—Briefing on Threat 

Environment Assessment (Closed—

Ex. 1). 

Week of December 19, 2005—Tentative 

There are no meetings scheduled for 

the Week of December 19, 2005. 

Week of December 26, 2005—Tentative 

There are no meetings scheduled for 

the Week of December 26, 2005. 

Week of January 2, 2006—Tentative 

There are no meetings scheduled for 

the Week of January 2, 2006. 

Week of January 9, 2006—Tentative 

Tuesday, January 10, 2006. 

9:30 a.m.—Briefing on International 

Research and Bilateral Agreements, 

(Contact: Roman Schaffer, 301–415–

7606). 

This meeting will be webcast live at 

the Web address: http://www.nrc.gov. 

Wednesday, January 11, 2006. 

9:30 a.m.—Meeting with Advisory 

Committee on Nuclear Waste 

(ACNW), (Contact: John Larkins, 

301–415–7360). 

This meeting will be webcast live at 

the Web address: http://www.nrc.gov. 

Thursday, January 12, 2006. 

9:30 a.m.—Discussion of Security 

Issues (Closed—Ex. 1 & 2). 

*The schedule for commission 

meetings is subject to change on short 

notice. To verify the status of meetings 

call (recording)—(301) 415–1292. 

contact person for more information: 

Michelle Schroll, (301) 415–1662. 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: http://www.nrc.gov/what-we-do/ 

policy-making/schedule.html. 

Additional Information 

The Affirmation Session tentatively 

scheduled on November 30, 2005, at 

9:25 a.m. has been rescheduled 

tentatively on December 12, 2005, at 

8:50 a.m. 

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g., 

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

August Spector, at 301–415–7080, TDD: 

301–415–2100, or by e-mail at 

aks@nrc.gov. Determinations on 

requests for reasonable accommodation 

will be made on a case-by-case basis. 

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov. 

Dated: December 1, 2005. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 05–23706 Filed 12–2–05; 11:00 am] 

BILLING CODE 7590–01–M 

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 to 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 November 9, 

2005 to November 21, 2005. The last 

biweekly notice was published on 

November 22, 2005 (70 FR 70641). 

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 

72668 Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 

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, http://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 

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 email 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, http:// 

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–

Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72669 

4209, (301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Calvert Cliffs Nuclear Power Plant, Inc., 

Docket Nos. 50–317 and 50–318, Calvert 

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of amendments request: July 13, 

2005. 

Description of amendments request: 

The proposed amendment would revise 

Technical Specification (TS) 1.1, 

‘‘Definitions,’’ TS 3.4.13, ‘‘RCS [reactor 

coolant system] Operational Leakage,’’

TS 5.5.9, ‘‘Steam Generator Tube 

Surveillance Program,’’ and TS 5.6.9, 

‘‘Steam Generator Tube Inspection 

Report,’’ and add a new specification 

(TS 3.4.18) for Steam Generator (SG) 

Tube Integrity. The proposed changes 

are necessary in order to implement the 

guidance for the industry initiative on 

Nuclear Energy Institute (NEI) 97–06, 

‘‘Steam Generator Program Guidelines.’’

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on March 2, 2005 (70 FR 

10298), on possible amendments 

adopting Technical Specification Task 

Force Change Traveller 449, 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 6, 2005 (70 FR 24126). 

The licensee affirmed the applicability 

of the following NSHC determination in 

its application dated July 13, 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 requires a SG 

Program that includes performance criteria 

that will provide reasonable assurance that 

the SG tubing will retain integrity over the 

full range of operating conditions (including 

startup, operation in the power range, hot 

standby, cooldown and all anticipated 

transients included in the design 

specification). The SG performance criteria 

are based on tube structural integrity, 

accident induced leakage, and operational 

LEAKAGE. 

A SGTR [steam generator tube rupture] 

event is one of the design basis accidents that 

are analyzed as part of a plant’s licensing 

basis. In the analysis of a SGTR event, a 

bounding primary to secondary LEAKAGE 

rate equal to the operational LEAKAGE rate 

limits in the licensing basis plus the 

LEAKAGE rate associated with a doubleended rupture of a single tube is assumed. 

For other design basis accidents such as 

MSLB [main steam line break], rod ejection, 

and reactor coolant pump locked rotor the 

tubes are assumed to retain their structural 

integrity (i.e., they are assumed not to 

rupture). These analyses typically assume 

that primary to secondary LEAKAGE for all 

SGs is 1 gallon per minute or increases to 1 

gallon per minute as a result of accident 

induced stresses. The accident induced 

leakage criterion introduced by the proposed 

changes accounts for tubes that may leak 

during design basis accidents. The accident 

induced leakage criterion limits this leakage 

to no more than the value assumed in the 

accident analysis. 

The SG performance criteria proposed 

change to the TS identify the standards 

against which tube integrity is to be 

measured. Meeting the performance criteria 

provides reasonable assurance that the SG 

tubing will remain capable of fulfilling its 

specific safety function of maintaining 

reactor coolant pressure boundary integrity 

throughout each operating cycle and in the 

unlikely event of a design basis accident. The 

performance criteria are only a part of the SG 

Program required by the proposed change to 

the TS. The program, defined by NEI 97–06, 

Steam Generator Program Guidelines, 

includes a framework that incorporates a 

balance of prevention, inspection, evaluation, 

repair, and leakage monitoring. The proposed 

changes do not, therefore, significantly 

increase the probability of an accident 

previously evaluated. 

The consequences of design basis accidents 

are, in part, functions of the DOSE 

EQUIVALENT I–131 in the primary coolant 

and the primary to secondary LEAKAGE 

rates resulting from an accident. Therefore, 

limits are included in the plant technical 

specifications for operational leakage and for 

DOSE EQUIVALENT I–131 in primary 

coolant to ensure the plant is operated within 

its analyzed condition. The typical analysis 

of the limiting design basis accident assumes 

that primary to secondary leak rate after the 

accident is 1 gallon per minute with no more 

than [500 gallons per day or 720 gallons per 

day] in any one SG, and that the reactor 

coolant activity levels of DOSE 

EQUIVALENT I–131 are at the TS values 

before the accident. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary coolant chemistry controls. The 

proposed approach updates the current TSs 

and enhances the requirements for SG 

inspections. The proposed change does not 

adversely impact any other previously 

evaluated design basis accident and is an 

improvement over the current TSs. 

Therefore, the proposed change does not 

affect the consequences of a SGTR accident 

and the probability of such an accident is 

reduced. In addition, the proposed changes 

do not affect the consequences of an MSLB, 

rod ejection, or a reactor coolant pump 

locked rotor event, or other previously 

evaluated accident. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident from any Previously 

Evaluated 

The proposed performance based 

requirements are an improvement over the 

requirements imposed by the current 

technical specifications. Implementation of 

the proposed SG Program will not introduce 

any adverse changes to the plant design basis 

or postulated accidents resulting from 

potential tube degradation. The result of the 

implementation of the SG Program will be an 

enhancement of SG tube performance. 

Primary to secondary LEAKAGE that may be 

experienced during all plant conditions will 

be monitored to ensure it remains within 

current accident analysis assumptions. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary or secondary coolant chemistry 

controls. In addition, the proposed change 

does not impact any other plant system or 

component. The change enhances SG 

inspection requirements. 

Therefore, the proposed change does not 

create the possibility of a new or different 

[kind] of accident from any accident 

previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in [a] Margin 

of Safety 

The SG tubes in pressurized water reactors 

are an integral part of the reactor coolant 

pressure boundary and, as such, are relied 

upon to maintain the primary system’s 

pressure and inventory. As part of the reactor 

coolant pressure boundary, the SG tubes are 

unique in that they are also relied upon as 

a heat transfer surface between the primary 

and secondary systems such that residual 

heat can be removed from the primary 

system. In addition, the SG tubes isolate the 

radioactive fission products in the primary 

coolant from the secondary system. In 

summary, the safety function of an SG is 

maintained by ensuring the integrity of its 

tubes. 

Steam generator tube integrity is a function 

of the design, environment, and the physical 

condition of the tube. The proposed change 

does not affect tube design or operating 

environment. The proposed change is 

expected to result in an improvement in the 

tube integrity by implementing the SG 

Program to manage SG tube inspection, 

assessment, repair, and plugging. The 

requirements established by the SG Program 

are consistent with those in the applicable 

design codes and standards and are an 

improvement over the requirements in the 

current TSs. 

For the above reasons, the margin of safety 

is not changed and overall plant safety will 

be enhanced by the proposed change to the 

TS. 

The NRC staff proposes to determine 

that the amendments request involves 

no significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Sr. Counsel—Nuclear Generation, 

Constellation Generation Group, LLC, 

750 East Pratt Street, 17th floor, 

Baltimore, MD 21202. 

72670 Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 

NRC Branch Chief: Richard J. Laufer. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of amendment request: October 

31, 2005. 

Description of amendment request: 

The proposed amendment change 

would add Technical Specification (TS) 

Limiting Condition for Operation (LCO) 

3.0.8, to allow 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 consistent with the 

program in place for complying with the 

requirements of 10 CFR 50.65(a)(4). In 

addition, a proposed change to LCO 

3.0.1 is required to reference the 

addition of LCO 3.0.8. 

This change was proposed by the 

industry’s Technical Specification Task 

Force (TSTF) and is designated as 

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 October 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 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. The postulated 

seismic event requiring snubbers is a lowprobability 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 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 1.177. A bounding risk assessment was 

performed to justify the proposed TS 

changes. This application of LCO 3.0.8 is 

predicated upon the licensee’s performance 

of a risk assessment and the management of 

plant risk. The net change to the margin of 

safety is insignificant. Therefore, this change 

does not involve a significant reduction in a 

margin of safety. 

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David G. 

Pettinari, Legal Department, 688 WCB, 

Detroit Edison Company, 2000 2nd 

Avenue, Detroit, Michigan 48226–1279. 

NRC Branch Chief: L. Raghavan. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 3, 

Westchester County, New York 

Date of amendment request: October 

3, 2005. 

Description of amendment request: 

The proposed amendment revises 

Technical Specification (TS) 

Surveillance Requirements (SRs) to 

reflect changes to the Emergency Core 

Cooling System throttle valves. The 

proposed amendment will add the 

modified throttle valves to the 

surveillance, remove existing throttle 

valves that are now locked closed from 

the surveillance, and add existing valves 

to the surveillance that are used in a 

throttle position when open. 

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 Surveillance 

Requirement (SR) 3.5.2.6 adds nine valves 

and removes two valves in the High Head 

Safety Injection (HHSI) system discharge 

lines. The SR requires verification that 

identified ECCS [emergency core cooling 

system] throttle valves position stops are in 

the correct position. The change reflects a 

stretch power uprate (SPU) modification that 

added throttle valves SI–2165, 2166, 2168, 

2169, 2170, 2171, and 2172, and locked 

closed valves Sl-856A and 856F. This 

amendment is adding to the SR those throttle 

valves which are now under administrative 

control and deletes the valves which no 

longer perform a throttle function. The 

amendment also adds hot leg valves Sl-856B 

and 856G which are used as throttle valves 

but never included in the SR. Valve Sl-856G 

still performs a throttle function and valve 

SI–856B can still be considered a throttle 

valve when used to trim system resistance. 

Verification of valve position has no effect on 

the probability of an accident previously 

evaluated since the HHSI system is not 

associated with the initiation of any accident. 

The verification of valve positions that will 

be required by the revised SR provides 

additional assurance that the HHSI throttle 

valves are in the position that is established 

by flow testing. Providing assurance of 

required valve positions does not increase 

the consequences of an accident previously 

evaluated.Therefore, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

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 to Surveillance 

Requirement 3.5.2.6 adds nine valves and 

removes two valves in the High Head Safety 

Injection (HHSI) system discharge lines. The 

SR requires verification that identified ECCS 

throttle valves position stops are in the 

correct position. The change corrects a 

deficient surveillance and does not affect the 

function of the valves or otherwise affect the 

design and operation of plant systems and 

components and therefore no new accident 

Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72671 

scenarios would be created. Therefore, no 

new failure modes are being introduced that 

could lead to different accidents. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change to Surveillance 

Requirement 3.5.2.6 adds nine valves and 

removes two valves in the High Head Safety 

Injection (HHSI) system discharge lines. The 

SR requires verification that identified ECCS 

throttle valves position stops are in the 

correct position. The change reflects a stretch 

power uprate (SPU) modification that added 

throttle valves SI–2165, 2166, 2168, 2169, 

2170, 2171, and 2172, and locked closed 

valves Sl-856A and 856F. The proposed 

amendment also adds valves SI–856B and 

856G which are used as throttle valves but 

never included in the SR. Valve Sl-856G still 

performs a throttle function and valve Sl856B can still be considered a throttle valve 

when used to trim system resistance. The 

frequency for verification of throttle valve 

stop positions is not altered by this 

amendment so this has no effect on the 

margin of safety. The valves for which 

verification of positions stops is required 

reflect the manner in which the system is 

currently analyzed and configured so the 

proposed change serves to maintain the 

required margin of safety by adding to the 

Technical Specifications the surveillances 

presently being administratively controlled. 

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: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Branch Chief: Richard J. Laufer. 

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of amendment request: June 29, 

2005. 

Description of amendment request: 

Entergy Operations, Incorporated 

(Entergy) proposes to relocate the 

License Condition associated with the 

Shutdown Cooling (SDC) Open 

Permissive Interlock (OPI) to the 

Technical Requirements Manual (TRM). 

The Nuclear Regulatory Commission 

(NRC) approved Standard Technical 

Specifications, Combustion Engineering 

Plants (NUREG–1432) include a 

surveillance requirement for this 

function due to the complexity and 

differences of plant designs, which 

would not support complete removal of 

the function from the NUREG. For 

Arkansas Nuclear One, Unit 2 (ANO–2), 

however, the OPI is not an assumed 

function that supports the accident 

analysis and does not meet the criteria 

in Section 50.36 of Title 10 of the Code 

of Federal Regulations (10 CFR) for 

inclusion in the technical specifications. 

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 OPI function is not required to ensure 

continued safe operation of the ANO–2 

facility. The OPI function is not considered 

an accident precursor or relied upon as a 

means of accident mitigation. The proposed 

change has no affect on plant design or 

operation. 

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 relocation of the OPI function to the 

TRM does not require any physical alteration 

to the plant or alter plant design. The OPI 

function is not considered an accident 

initiator nor is this function credited in any 

safety analyses for the prevention or 

mitigation of any accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The OPI function is not credited in a 

margin of safety analysis for any accident 

previously evaluated. Relocation of the OPI 

function requirements will not result in a 

credible increase in nuclear safety risk. 

Appropriate alarm, design features, and 

administrative controls continue to ensure 

proper isolation of the SDC system during 

plant operations with elevated RCS [reactor 

cooling system] pressures. In addition, the 

OPI function will be relocated to the TRM, 

which is part of the Safety Analysis Report 

(SAR) and controlled by 10 CFR 50.59. 

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: Nicholas S. 

Reynolds, Esquire, Winston and 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 No. 2, 

Pope County, Arkansas 

Date of amendment request: 

September 19, 2005. 

Description of amendment request: 

The proposed change will modify the 

Surveillance Requirements related to 

Arkansas One, Unit 2, technical 

specification (TS) 3.1.1.4, Moderator 

Temperature Coefficient (MTC), and 

will allow the use of topical report 

WCAP–16011-P-A, ‘‘Startup Test 

Activity Reduction Program.’’ A change 

to NUREG–1432, ‘‘Standard Technical 

Specifications Combustion Engineering 

Plants,’’ has been proposed in Technical 

Specification Task Force (TSTF) 

Improved Standard Technical 

Specification Change Traveler TSTF–

486 to incorporate the allowance to use 

WCAP–16011–P–A. The traveler was 

submitted for Nuclear Regulatory 

Commission (NRC) approval in June 

2005. 

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 MTC is not an initiator of any 

previously evaluated accidents. As an input 

into accident analyses, the MTC is used to 

predict plant behavior in the event of an 

accident. It was demonstrated in WCAP–

16011-P-A that the modified MTC 

verification (i.e., measured RCS [reactor 

coolant system] boron concentration) is 

adequate to ensure that the MTC remains 

within the limits provided the STAR 

applicability requirements are met. It was 

also demonstrated in WCAP–16011-P-A that 

the elimination of the EOC [emergency 

operations center] MTC measurement is 

acceptable when the applicability 

requirements given in WCAP–16011-P-A are 

met and the result of the MTC determination 

performed prior to reaching a Rated Thermal 

Power equilibrium boron concentration of 

800 ppm is within a tolerance of ± 0.16 × 

10¥4 Dk/k/°F from the corresponding design 

value. 

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 

72672 Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of structure, system, or 

component will be installed). 

The methods governing normal plant 

operations are not altered 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 a margin of safety? 

Response: No. 

The proposed change will not affect the 

margin of safety. The MTC limits are 

unaffected and an acceptable method will be 

used to demonstrate that MTC is within its 

limits. 

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: Nicholas S. 

Reynolds, Esquire, Winston and 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 No. 2 

(ANO–2), Pope County, Arkansas 

Date of amendment request: 

September 19, 2005. 

Description of amendment request: 

The proposed change will modify the 

ANO–2 technical specification (TS) 

3.1.1.5, Minimum Temperature for 

Criticality. Specifically, the proposed 

change will raise the minimum 

temperature for criticality from the 

current value of 3 525 °F to 3 540 °F. 

Changes are also proposed to the Action 

statement and Surveillance 

Requirements to support the increase in 

temperature. The change is needed to 

support core design. 

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. 

There are no accident analyses that dictate 

the minimum temperature for criticality. The 

minimum temperature for criticality is not an 

accident initiator. It is used in the reload 

analyses as a limiting temperature at which 

the core design is verified to satisfy the limit 

of the positive moderator temperature 

coefficient (MTC) specified in the ANO–2 TS 

and Core Operating Limits Report (COLR). 

The minimum temperature for criticality is 

one of many input parameters used in the 

reload design analytical calculation that 

confirms the core design satisfies the MTC 

TS and COLR. 

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 to increase the 

minimum temperature for criticality does not 

result in any plant design changes. In 

addition, the minimum temperature at which 

the reactor is taken critical is not an accident 

initiator. The nominal average reactor coolant 

system temperature during an approach to 

criticality is several degrees higher than the 

limit proposed for the minimum temperature 

for criticality. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The increase of the minimum temperature 

for criticality in conjunction with the use of 

a sufficient number of burnable absorber 

rods, which will be incorporated into the 

core design, will ensure the current TS 

limits, as reflected in the COLR, for the most 

positive MTC will continue to be satisfied. 

The current transient analysis results are 

bounding and remain applicable. 

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: Nicholas S. 

Reynolds, Esquire, Winston and Strawn, 

1700 K Street, NW., Washington, DC 

20006–3817. 

NRC Branch Chief: David Terao. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: October 

25, 2005. 

Description of amendment request: 

The proposed change will modify the 

Waterford 3 Technical Specification 

(TS) 3.1.1.4, Minimum Temperature for 

Criticality. Specifically, the proposed 

change will raise the minimum 

temperature for criticality from the 

current value of ≥520°F to ≥533°F. 

Changes are also proposed to the Action 

statement and Surveillance 

Requirements to support the increase in 

temperature. 

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 minimum temperature for criticality is 

not an accident initiator. It is used in the 

reload analyses as a limiting temperature at 

which the core design is verified to satisfy 

the limit of the positive moderator 

temperature coefficient (MTC) specified in 

the Waterford 3 TS and Core Operating 

Limits Report (COLR). The minimum 

temperature for criticality is one of many 

input parameters used in the reload design 

analytical calculation that confirms the core 

design satisfies the MTC TS and COLR. 

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 to increase the 

minimum temperature for criticality does not 

result in any plant design changes. In 

addition the minimum temperature at which 

the reactor is taken critical is not an accident 

initiator. The nominal average reactor coolant 

system temperature during an approach to 

criticality is several degrees higher than the 

limit proposed for the minimum temperature 

for criticality. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The increase of the minimum temperature 

for criticality in conjunction with the 

appropriate core designs will ensure the 

current TS limits, as reflected in the COLR, 

for the most positive MTC will continue to 

be satisfied. 

The current transient analysis results are 

bounding and remain applicable. 

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 

Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72673 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: N. S. Reynolds, 

Esquire, Winston & Strawn, 1700 K 

Street NW., Washington, DC 20006–

3817. 

NRC Branch Chief: David Terao. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: October 

25, 2005. 

Description of amendment request: 

The proposed change will modify the 

Surveillance Requirements (SRs) related 

to Waterford 3 Technical Specification 

(TS) 3.1.1.3, Moderator Temperature 

Coefficient (MTC) and will allow the 

use of the Startup Test Activity 

Reduction Program (WCAP–16011–P–

A). 

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 MTC is not an initiator of any 

previously evaluated accidents. As an input 

into accident analyses, the MTC is used to 

predict plant behavior in the event of an 

accident. It was demonstrated in WCAP–

16011–P–A that the modified MTC 

verification (i.e., measured RCS [reactor 

coolant system] boron concentration) is 

adequate to ensure that the MTC remains 

within the limits, provided the STAR 

applicability requirements are met. It was 

also demonstrated in WCAP–16011–P–A that 

the elimination of the EOC [end-of-cycle] 

MTC measurement is acceptable when the 

applicability requirements given in WCAP–

16011–P–A are met and the result of the MTC 

determination performed at greater than 15 

percent of Rated Thermal Power and prior to 

reaching 40 EFPD [effective full power days] 

is within a tolerance of ± 0.16 × 10¥4 Dk/k/ 

°F from the corresponding design value. 

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 does not involve a 

physical alteration to the plant (i.e., no new 

or different type of structure, system, or 

component will be installed). The methods 

governing normal plant operations are not 

altered by the proposed TS 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 a margin of safety? 

Response: No. 

The proposed change will not affect the 

margin of safety. The MTC limits are 

unaffected and an acceptable method will be 

used to demonstrate that MTC is within its 

limits. 

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: N. S. Reynolds, 

Esquire, Winston & Strawn, 1700 K 

Street NW., Washington, DC 20006–

3817. 

NRC Branch Chief: David Terao. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: October 

25, 2005. 

Description of amendment request: 

The proposed change to Technical 

Specification 6.9.1.11, Core Operating 

Limits Report, will result in the addition 

of a methodology that will allow the use 

of zirconium diboride (ZrB2) burnable 

absorber coating on fuel pellets. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change will add topical 

report WCAP–16072–P–A to the NRC 

reviewed and approved analytical methods 

used to determine the core operating limits. 

The topical report has been previously 

approved by the NRC for use in Combustion 

Engineering core designs and as such, the 

proposed change is administrative in nature 

and has no impact on any plant 

configurations or on system performance that 

is relied upon to mitigate the consequences 

of an accident. In addition, prior to the use 

of the ZrB2 burnable absorber coating, fuel 

design will be analyzed with applicable NRC 

staff approved codes and methods. 

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 adds a reference to 

the topical report that allows the use of ZrB2

as a burnable absorber coating on the fuel 

pellet. The topical report has been previously 

approved by the NRC for use in Combustion 

Engineering core designs and as such, the 

proposed change is administrative in nature 

and has no impact on any plant 

configurations or on system performance that 

is relied upon to mitigate the consequences 

of an accident. In addition, prior to the use 

of the ZrB2 burnable absorber coating, fuel 

design will be analyzed with applicable NRC 

staff approved codes and methods. This 

change is administrative in nature and does 

not create a new or different type of accident 

than previously evaluated because the design 

requirements for the facility remain the same. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change will add WCAP–

16072–P–A to the list of referenced topical 

reports. The topical report has been 

previously approved by the NRC for use in 

Combustion Engineering core designs and as 

such, the proposed change is administrative 

in nature and has no impact on any plant 

configurations or on system performance that 

is relied upon to mitigate the consequences 

of an accident. In addition, prior to the use 

of the ZrB2 burnable absorber coating, fuel 

design will be analyzed with applicable NRC 

staff approved codes and methods. 

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: N. S. Reynolds, 

Esquire, Winston & Strawn, 1700 K 

Street NW., Washington, DC 20006–

3817. 

NRC Branch Chief: David Terao. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant, Units 1 and 2, 

Berrien County, Michigan 

Date of amendment requests: July 29, 

2005. 

Description of amendment requests: 

The proposed amendments would 

delete requirements from the Technical 

Specifications (TSs) to submit monthly 

operating reports and annual 

occupational radiation exposure reports. 

The changes are consistent with 

72674 Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 

Revision 1 of Nuclear Regulatory 

Commission (NRC) approved Industry/ 

Technical Specifications Task Force 

(TSTF) Standard Technical 

Specification Change Traveler, TSTF–

369, ‘‘Removal of Monthly Operating 

and Occupational Radiation Exposure 

Report.’’ The availability of this TS 

improvement was announced in the 

Federal Register (69 FR 35067) on June 

23, 2004, as part of the Consolidated 

Line Item Improvement Process (CLIIP). 

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in license 

amendment applications in the Federal 

Register on April 29, 2004 (69 FR 

23542). The licensee affirmed the 

applicability of the model NSHC 

determination in its application dated 

July 29, 2005. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC (which 

was previously published in 69 FR 

23542) 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 eliminates the 

Technical Specifications reporting 

requirements to provide a monthly operating 

report of shutdown experience and operating 

statistics if the equivalent data is submitted 

using an industry electronic database. It also 

eliminates the Technical Specification 

reporting requirement for an annual 

occupational radiation exposure report, 

which provides information beyond that 

specified in NRC regulations. The proposed 

change involves no changes to plant systems 

or accident analyses. As such, the change is 

administrative in nature and does not affect 

initiators of analyzed events or assumed 

mitigation of accidents or transients. 

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 does not involve a 

physical alteration of the plant, add any new 

equipment, or require any existing 

equipment to be operated in a manner 

different from the present design. 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. 

This is an administrative change to 

reporting requirements of plant operating 

information and occupational radiation 

exposure data, and has no effect on plant 

equipment, operating practices or safety 

analyses assumptions. For these reasons, the 

proposed change does not involve a 

significant reduction in the margin of safety. 

Based on the reasoning presented 

above and the previous discussion of 

the amendment request, the NRC staff 

proposes to determine that the 

requested change does not involve a 

significant hazards consideration. 

Attorney for licensee: James M. Petro, 

Jr., Esquire, One Cook Place, Bridgman, 

MI 49106. 

NRC Branch Chief: L. Raghavan. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant, Units 1 and 2, 

Berrien County, Michigan 

Date of amendment requests: August 

10, 2005. 

Description of amendment requests: 

The proposed amendments would 

delete the power range neutron flux 

high negative rate trip function from 

each unit’s Technical Specifications. 

The licensee’s proposed changes are 

based on the methodology presented in 

Westinghouse Topical Report WCAP–

11394–P–A, ‘‘Methodology for the 

Analysis of the Dropped Rod Event,’’

which had been previously accepted by 

the Nuclear Regulatory Commission 

staff. 

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 of 

occurrence or consequences of an accident 

previously evaluated? 

Response: No. 

The removal of the power range neutron 

flux high negative rate trip function from 

technical specifications does not increase the 

probability or consequences of reactor core 

damage accidents resulting from dropped 

Rod Cluster Control Assembly (RCCA) events 

previously analyzed. The safety functions of 

other safety-related systems and components, 

which are related to mitigation of these 

events, [will] not [be] altered. All other 

Reactor Trip System and Engineered Safety 

Features Actuation Systems protection 

functions are not impacted by the 

elimination of the trip function. The dropped 

RCCA accident analysis does not rely on the 

negative flux rate trip to safely shut down the 

plant. The safety analysis of the plant is 

unaffected by the proposed change. Since the 

safety analysis is unaffected, the calculated 

radiological releases associated with the 

analysis are not affected. 

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 does not adversely 

alter the design assumptions, conditions, or 

configuration of the facility or the manner in 

which the plant is operated. No new accident 

scenarios, failure mechanisms, or limiting 

single failures are introduced as a result of 

the proposed change. The proposed change 

does not challenge the performance or 

integrity of any safety-related systems or 

components. Nuclear Regulatory Commission 

(NRC)-approved Westinghouse Topical 

Report WCAP–11394–P–A, ‘‘Methodology for 

the Analysis of the Dropped Rod Event,’’

dated January 1990 has demonstrated that the 

negative flux rate trip function can be 

eliminated. 

Therefore, the proposed changes does not 

created the possibility of a new or different 

kind of accident from any previously 

evaluated. 

(3) Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margin of safety associated with the 

acceptance criteria of any accident is 

unchanged. It has been demonstrated that the 

negative flux rate trip function can be 

eliminated by the NRC-approved 

methodology described in WCAP–11394–P–

A. Donald C. Cook Nuclear Plant cyclespecific analyses have confirmed that for a 

dropped RCCA(s) event, limits on departure 

from nucleate boiling are not exceeded by 

eliminating the negative flux rate trip. The 

proposed change will have no [e]ffect on the 

availability, operability, or performance of 

safety-related systems and components. 

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 requests involve no 

significant hazards consideration. 

Attorney for licensee: James M. Petro, 

Jr., Esquire, One Cook Place, Bridgman, 

MI 49106. 

NRC Branch Chief: L. Raghavan. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: August 

11, 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 Paragraph 50.65(a)(4) of 

Title 10 of the Code of Federal 

Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72675 

Regulations. Limiting Condition for 

Operation (LCO) 2.0.1(3) 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 

August 11, 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 technical 

specification (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 lowprobability 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 

[LCO 2.0.1(3) for Fort Calhoun Station] 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 [LCO 

2.0.1(3)]. 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 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. This application of 

LCO 3.0.8 is predicated upon the licensee’s 

performance of a risk assessment and the 

management of plant risk. [The proposed 

LCO 2.0.1(3) 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: James R. 

Curtiss, Esq., Winston & Strawn, 1400 L 

Street, NW., Washington, DC 20005–

3502. 

NRC Branch Chief: David Terao. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: 

November 8, 2005. 

Description of amendment request: 

The proposed amendment will modify 

Fort Calhoun Technical Specification 

(TS) 4.2.1, ‘‘Fuel Assemblies,’’ to permit 

the use of AREVA (Framatome ANP) 

M5TM advanced alloy for fuel rod 

cladding and structural components 

such as guide tubes, intermediate spacer 

grids, end plugs, and guide thimble 

tubes, beginning with Cycle 24. In 

addition, Omaha Public Power District 

proposes to modify TS 5.9 to include 

the Framatome ANP Topical Report 

evaluating the impact of M5TM material 

properties on NRC-approved 

methodology. M5TM is a proprietary, 

zirconium-based alloy that is a variant 

of Zr1Nb to replace zircaloy-4 in the 

construction of fuel assembly 

components. 

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 NRC[-]approved topical report BAW–

10[2]27P–A (Reference 8.1 [of amendment 

request]) that provides the licensing basis for 

M5TM cladding and structural material, has 

shown that the M5TM alloy exhibits superior 

properties to the currently used zircaloy-4 

material. The cladding by itself does not 

initiate an accident and therefore does not 

affect accident probability. It has been 

determined that M5TM cladding will not 

significantly affect the consequences of an 

accident. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not involve a significant increase in 

the probability or consequences of an 

accident previously analyzed. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not result in 

changes in the operation or overall 

configuration of the facility. Topical report 

BAW–10227P–A (Reference 8.1) 

demonstrated that the M5TM alloy will 

perform similar to or better than zircaloy-4, 

thus precluding the possibility of the fuel 

becoming an accident initiator and causing a 

new or different type of accident. 

Since the material properties of M5TM alloy 

are similar to or better than zircaloy-4, there 

will not be any significant change in the 

types of effluents that may be released offsite. There will not be any significant 

increase in occupational or public radiation 

exposure. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

AREVA has performed generic LOCA [lossof-coolant accident] and non-LOCA 

evaluations and demonstrated the use of the 

M5TM material will have only a small, or 

beneficial, impact on the event 

consequences. 

Plant-specific analyses using NRCapproved methodology for the mixed core 

will demonstrate that the reactor core safety 

limits will continue to be met. 

72676 Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 

Therefore, the proposed amendment 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: James R. 

Curtiss, Esq., Winston & Strawn, 1400 L 

Street, NW., Washington, DC 20005–

3502. 

NRC Branch Chief: David Terao. 

Southern California Edison Company, et 

al., Docket Nos. 50–361 and 50–362, 

San Onofre Nuclear Generating Station, 

Units 2 and 3, San Diego County, 

California 

Date of amendment requests: 

November 3, 2005. 

Description of amendment requests: 

The proposed amendment revises 

Technical Specification (TS) Section 

5.5.2.11 to modify the definitions of 

steam generator tube ‘‘Repair Limit’’ and 

‘‘Tube Inspection.’’ The purpose of 

these changes is to define the extent of 

the required tube inspections and repair 

criteria within the tubesheet regions. 

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 revises the San 

Onofre [Nuclear Generating Station,] Units 2 

and 3 Technical Specifications (TS) by 

revising the definitions of steam generator 

‘‘Repair Limit’’ and ‘‘Tube Inspection[,]’’ as 

contained in TS items 5.5.2.11.f.1.f and 

5.5.2.11.f.1.h, respectively. This proposed 

change also adds words in the ‘‘Operability 

determination’’ requirement (item 

5.5.2.11.f.2) to provide consistency with the 

proposed change in the definition of ‘‘Repair 

Limit.’’ These revisions maintain existing 

design limits and would not increase the 

probability or consequences of an accident 

involving tube burst or primary to secondary 

accident-induced leakage, as previously 

analyzed in the San Onofre [Nuclear 

Generating Station,] Units 2 and 3 Updated 

Final Safety Analysis Report (UFSAR). Also, 

the NEI 97–06 steam generator tube 

performance criterion for structural integrity 

and accident-induced leakage will continue 

to be satisfied. 

Tube burst is precluded for a tube with 

defects within the tubesheet region because 

of the constraint provided by the tubesheet. 

As such, tube pullout resulting from the axial 

forces induced by primary to secondary 

differential pressures would be a prerequisite 

for tube burst to occur. An industry test 

program (WCAP–16208–P Revision 1), and 

follow-on San Onofre site-specific analysis 

(WCAP–16208–P Revision 1, Supplement 1) 

defined the non-degraded hot leg tube to 

tubesheet joint length and cold leg tube to 

tubesheet joint length required to preclude 

tube pullout and maintain acceptable 

primary to secondary accident-induced 

leakage, assuming that 100% [percent] of the 

steam generator tubes experienced complete 

circumferential separation (360 degree 

through wall crack) immediately below both 

the hot leg recommended inspection length 

(C*) and the cold leg C*. Any degradation 

below C* is shown by empirical test results 

and analyses to be acceptable, thereby 

precluding an event with consequences 

similar to a postulated tube rupture event. 

WCAP–16208–P Revision 1, with 

Supplement 1 includes a total 0.2 gpm 

[gallons per minute]/steam generator 

assumed value for primary to secondary 

accident-induced leakage. Inspection to the 

C* lengths will ensure that the postulated 

accident-induced leakage will remain below 

the current primary to secondary leakage 

assumption utilized in the UFSAR accident 

analyses (Chapter 15). 

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. 

Steam generator tube leakage and 

structural integrity will be maintained during 

all plant conditions upon implementation of 

the proposed inspection scope and repair 

limit changes to the San Onofre [Nuclear 

Generating Station,] Unit 2 and 3 Technical 

Specifications. These changes do not 

introduce any new mechanisms that might 

result in a different kind of accident from 

those previously evaluated. Even with the 

limiting circumstances of complete 

circumferential separation (360 degree 

through wall crack) of all of the tubes below 

the C* length, [a] tube pullout is precluded 

and leakage is predicted to be maintained 

within accident analysis assumptions. 

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. 

Operation with potential tube degradation 

below the C* inspection length within the 

tubesheet region of the steam generator 

tubing meets the intent of the inspection 

guidance of Regulatory Guide Number 1.83, 

Revision 1, titled Inservice Inspection of 

Pressurized Water Reactor Steam Generator 

Tubes, the requirements of General Design 

Criteria 14, 15, 31 and 32 of 10 CFR 50, and 

the recommendations of the Nuclear Energy 

Institute in NEI 97–06, titled Steam Generator 

Program Guidelines. 

The total leakage from an undetected flaw 

population below the C* inspection length 

under postulated accident conditions is 

accounted for to assure that it is within the 

bounds of the accident analysis assumptions. 

Adequate margin remains for other possible 

steam generator tube leak sources. 

The proposed changes also maintain the 

structural and accident-induced leakage 

integrity of the steam generator tubes as 

required by NEI 97–06. 

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 requests involve no 

significant hazards consideration. 

Attorney for licensee: Douglas K. 

Porter, Esquire, Southern California 

Edison Company,2244 Walnut Grove 

Avenue, Rosemead, California 91770. 

NRC Branch Chief: David Terao. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: 

November 3, 2005. 

Description of amendment request: 

The amendment would revise the 

Technical Specifications (TS) to adopt 

NRC-approved Revision 4 to Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–449, ‘‘Steam 

Generator Tube Integrity.’’ The 

proposed amendment includes changes 

to the TS definition of Leakage, TS 

3.4.13, ‘‘RCS [Reactor Coolant System] 

Operational Leakage,’’ TS 5.5.9, ‘‘Steam 

Generator (SG) Program,’’ TS 5.6.9, 

‘‘Steam Generator Tube Inspection 

Report,’’ and adds TS 3.4.17, ‘‘Steam 

Generator (SG) Tube Integrity.’’ The 

proposed changes are necessary in order 

to implement the guidance for the 

industry initiative on NEI 97–06, 

‘‘Steam Generator Program Guidelines.’’

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on March 2, 2005 (70 FR 

10298), on possible amendments 

adopting TSTF–449, 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 6, 2005 (70 FR 24126). 

The licensee affirmed the applicability 

of the following NSHC determination in 

its application dated November 3, 2005. 

Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72677 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change requires an SG 

Program that includes performance criteria 

that will provide reasonable assurance that 

the SG tubing will retain integrity over the 

full range of operating conditions (including 

startup, operation in the power range, hot 

standby, cooldown and all anticipated 

transients included in the design 

specification). The SG performance criteria 

are based on tube structural integrity, 

accident induced leakage, and operational 

LEAKAGE. 

A steam generator tube rupture (SGTR) 

event is one of the design-basis accidents that 

are analyzed as part of a plant’s licensing 

basis. In the analysis of an SGTR event, a 

bounding primary to secondary LEAKAGE 

rate equal to the operational LEAKAGE rate 

limits in the licensing basis plus the 

LEAKAGE rate associated with a doubleended rupture of a single tube is assumed. 

For other design-basis accidents such as a 

main steamline break (MSLB), rod ejection, 

and reactor coolant pump locked rotor, the 

tubes are assumed to retain their structural 

integrity (i.e., they are assumed not to 

rupture). These analyses typically assume 

that primary to secondary LEAKAGE for all 

SGs are 1 gallon per minute or increases to 

1 gallon per minute as a result of accidentinduced stresses. The accident-induced 

leakage criterion introduced by the proposed 

changes accounts for tubes that may leak 

during design-basis accidents. The accidentinduced leakage criterion limits this leakage 

to no more than the value assumed in the 

accident analysis. 

The SG performance criteria proposed 

change to the TS identify the standards 

against which tube integrity is to be 

measured. Meeting the performance criteria 

provides reasonable assurance that the SG 

tubing will remain capable of fulfilling its 

specific safety function of maintaining 

reactor coolant pressure boundary integrity 

throughout each operating cycle and in the 

unlikely event of a design-basis accident. The 

performance criteria are only a part of the SG 

Program required by the proposed change to 

the TS. The program, defined by NEI 97–06, 

Steam Generator Program Guidelines, 

includes a framework that incorporates a 

balance of prevention, inspection, evaluation, 

repair, and leakage monitoring. The proposed 

changes do not, therefore, significantly 

increase the probability of an accident 

previously evaluated. 

The consequences of design-basis 

accidents are, in part, functions of the DOSE 

EQUIVALENT I–131 in the primary coolant 

and the primary to secondary LEAKAGE 

rates resulting from an accident. Therefore, 

limits are included in the plant technical 

specifications for operational leakage and for 

DOSE EQUIVALENT I–131 in primary 

coolant to ensure the plant is operated within 

its analyzed condition. The typical analysis 

of the limiting design-basis accident assumes 

that primary to secondary leak rate after the 

accident is 1 gallon per minute with no more 

than [500 gallons per day or 720 gallons per 

day] in any one SG, and that the reactor 

coolant activity levels of DOSE 

EQUIVALENT I–131 are at the TS values 

before the accident. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary coolant chemistry controls. The 

proposed approach updates the current TSs 

and enhances the requirements for SG 

inspections. The proposed change does not 

adversely impact any other previously 

evaluated design-basis accident and is an 

improvement over the current TSs. 

Therefore, the proposed change does not 

affect the consequences of an SGTR accident, 

and the probability of such an accident is 

reduced. In addition, the proposed changes 

do not affect the consequences of an MSLB, 

rod ejection, or a reactor coolant pump 

locked rotor event, or other previously 

evaluated accident. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Previously 

Evaluated 

The proposed performance-based 

requirements are an improvement over the 

requirements imposed by the current 

technical specifications. Implementation of 

the proposed SG Program will not introduce 

any adverse changes to the plant design basis 

or postulated accidents resulting from 

potential tube degradation. The result of the 

implementation of the SG Program will be an 

enhancement of SG tube performance. 

Primary to secondary LEAKAGE that may be 

experienced during all plant conditions will 

be monitored to ensure it remains within 

current accident analysis assumptions. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary or secondary coolant chemistry 

controls. In addition, the proposed change 

does not impact any other plant system or 

component. The change enhances SG 

inspection requirements. 

Therefore, the proposed change does not 

create the possibility of a new or different 

type of accident from any accident 

previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The SG tubes in pressurized-water reactors 

are an integral part of the reactor coolant 

pressure boundary and, as such, are relied 

upon to maintain the primary system’s 

pressure and inventory. As part of the reactor 

coolant pressure boundary, the SG tubes are 

unique in that they are also relied upon as 

a heat transfer surface between the primary 

and secondary systems such that residual 

heat can be removed from the primary 

system. In addition, the SG tubes isolate the 

radioactive fission products in the primary 

coolant from the secondary system. In 

summary, the safety function of an SG is 

maintained by ensuring the integrity of its 

tubes. 

Steam generator tube integrity is a function 

of the design, environment, and the physical 

condition of the tube. The proposed change 

does not affect tube design or operating 

environment. The proposed change is 

expected to result in an improvement in the 

tube integrity by implementing the SG 

Program to manage SG tube inspection, 

assessment, repair, and plugging. The 

requirements established by the SG Program 

are consistent with those in the applicable 

design codes and standards and are an 

improvement over the requirements in the 

current TSs. 

For the above reasons, the margin of safety 

is not changed and overall plant safety will 

be enhanced by the proposed change to the 

TS. 

The NRC staff proposes to determine 

that the amendments request involves 

no significant hazards consideration. 

Attorney for licensee: Jay Silberg, Esq., 

Shaw, Pittman, Potts and Trowbridge, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: David Terao. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

of 1954, as amended (the Act), and the 

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

amendment, (2) the amendment, and (3) 

the Commission’s related letter, Safety 

72678 Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (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 

Systems (ADAMS) Public Electronic 

Reading Room on the internet at the 

NRC web site, http://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. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendments: 

July 9, 2004. 

Brief description of amendments: The 

amendments revise the Operating 

Licenses and Technical Specifications 

(TSs) to allow operation of Palo Verde 

Nuclear Generating Station (PVNGS), 

Units 1 and 3 up to a maximum reactor 

core power level of 3990 Megawatts 

thermal (MWt), an increase of 2.94 

percent above the current licensed 

power level of 3876 MWt. The proposed 

amendments would also make 

administrative changes to the PVNGS 

Unit 2 TSs so that the changed pages 

would apply to the three PVNGS units. 

Operation at the uprated power level 

with replacement steam generators has 

been approved for PVNGS Unit 2. 

Date of issuance: November 16, 2005. 

Effective date: November 16, 2005, 

and shall be implemented within 90 

days of the date of issuance. 

Amendment Nos.: Unit 1–157, Unit 

2–157, Unit 3–157. 

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The 

amendments revise the Operating 

Licenses for Units 1 and 3 and the 

Technical Specifications for all three 

units. 

Date of initial notice in Federal 

Register: September 28, 2004 (69 FR 

57980). The June 2, June 3 (two letters), 

June 17, July 9 (two letters), July 19 (two 

letters), August 3, September 29, 

October 21, and November 1, 2005, 

supplemental letters provided 

additional clarifying information, did 

not expand the scope of the application 

as originally noticed, and did not 

change the NRC staff’s original proposed 

no significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 16, 

2005. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States, Inc., and Entergy 

Operations, Inc., Docket No. 50–458, 

River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: August 

31, 2005, as supplemented by letter 

dated September 13, 2005. 

Brief description of amendment: The 

amendment permitted a one-time 

change to Technical Specification Table 

3.3.8.1–1 to provide a one-time 

relaxation of the Loss of Power 

instrumentation requirements. 

Date of issuance: September 15, 2005. 

Effective date: As of the date of 

issuance to be implemented 

immediately. 

Amendment No.: 147. 

Facility Operating License No. NPF–

47: Amendment revised the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration: Yes. The NRC published 

a public notice of the proposed 

amendment, issued a proposed finding 

of no significant hazards consideration, 

and requested that any comments on the 

proposed no significant hazards 

consideration be provided to the NRC 

staff by the close of business on 

September 9, 2005. The notice was 

published in The St. Francisville 

Democrat (in St. Francisville) on 

September 8, 2005, and The Advocate 

(in Baton Rouge) on September 7, 2005. 

No public comments were received. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, consultation with the 

State of Louisiana, and final no 

significant hazards consideration 

determination are contained in a Safety 

Evaluation dated September 15, 2005. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of application for amendment: 

November 1, 2004, as supplemented by 

letters dated April 12, July 22, and 

September 26, 2005. 

Brief description of amendment: The 

amendment authorizes the use of a 

single-failure-proof gantry crane for 

spent fuel cask handling operations up 

to 110 tons in weight. 

Date of issuance: November 21, 2005. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 244. 

Facility Operating License No. DPR–

26: The amendment allows use of the 

gantry crane for spent fuel cask 

handling operations up to 110 tons in 

weight. 

Date of initial notice in Federal 

Register: December 7, 2004 (69 FR 

70716). The April 12, July 22, and 

September 26, 2005, supplements 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the NRC 

staff’s original proposed no significant 

hazards consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 21, 

2005. 

No significant hazards consideration 

comments received: No. 

Pacific Gas and Electric Company, 

Docket Nos. 50–275 and 50–323, Diablo 

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments: 

November 3, 2004, and its supplements 

dated February 24, June 23, and 

September 30, 2005. 

Brief description of amendments: The 

amendments allow installation and use 

of a temporary cask pit spent fuel 

storage rack for Units 1 and 2. The cask 

pit rack would allow the storage of an 

additional 154 spent fuel assemblies for 

each unit. The total spent fuel pool 

storage capacity for each unit would be 

increased from the current 1324 spent 

fuel assemblies to 1478 assemblies for 

Cycles 14–16. 

Date of issuance: November 21, 2005. 

Effective date: As of the date of 

issuance, and shall be implemented 

upon installation of the temporary cask 

pit spent fuel rack. 

Amendment Nos.: Unit 1—183; Unit 

2–185. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: December 21, 2004 (69 FR 

76481). The February 24, June 23, and 

September 30, 2005, supplemental 

letters provided additional clarifying 

information, did not expand the scope 

of the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72679 

Safety Evaluation dated November 21, 

2005. 

No significant hazards consideration 

comments received: Yes. The comments 

are addressed in the enclosure of the 

above Safety Evaluation. 

Notice of Issuance of Amendments to 

Facility Operating Licenses and Final 

Determination of No Significant 

Hazards Consideration and 

Opportunity for a Hearing (Exigent 

Public Announcement or Emergency 

Circumstances) 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application for the 

amendment complies with the 

standards and requirements of the 

Atomic Energy Act of 1954, as amended 

(the Act), and the Commission’s rules 

and regulations. The Commission has 

made appropriate findings as required 

by the Act and the Commission’s rules 

and regulations in 10 CFR Chapter I, 

which are set forth in the license 

amendment. 

Because of exigent or emergency 

circumstances associated with the date 

the amendment was needed, there was 

not time for the Commission to publish, 

for public comment before issuance, its 

usual Notice of Consideration of 

Issuance of Amendment, Proposed No 

Significant Hazards Consideration 

Determination, and Opportunity for a 

Hearing. 

For exigent circumstances, the 

Commission has either issued a Federal 

Register notice providing opportunity 

for public comment or has used local 

media to provide notice to the public in 

the area surrounding a licensee’s facility 

of the licensee’s application and of the 

Commission’s proposed determination 

of no significant hazards consideration. 

The Commission has provided a 

reasonable opportunity for the public to 

comment, using its best efforts to make 

available to the public means of 

communication for the public to 

respond quickly, and in the case of 

telephone comments, the comments 

have been recorded or transcribed as 

appropriate and the licensee has been 

informed of the public comments. 

In circumstances where failure to act 

in a timely way would have resulted, for 

example, in derating or shutdown of a 

nuclear power plant or in prevention of 

either resumption of operation or of 

increase in power output up to the 

plant’s licensed power level, the 

Commission may not have had an 

opportunity to provide for public 

comment on its no significant hazards 

consideration determination. In such 

case, the license amendment has been 

issued without opportunity for 

comment. If there has been some time 

for public comment but less than 30 

days, the Commission may provide an 

opportunity for public comment. If 

comments have been requested, it is so 

stated. In either event, the State has 

been consulted by telephone whenever 

possible. 

Under its regulations, the Commission 

may issue and make an amendment 

immediately effective, notwithstanding 

the pendency before it of a request for 

a hearing from any person, in advance 

of the holding and completion of any 

required hearing, where it has 

determined that no significant hazards 

consideration is involved. 

The Commission has applied the 

standards of 10 CFR 50.92 and has made 

a final determination that the 

amendment involves no significant 

hazards consideration. The basis for this 

determination is contained in the 

documents related to this action. 

Accordingly, the amendments have 

been issued and made effective as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the application for 

amendment, (2) the amendment to 

Facility Operating License, and (3) the 

Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment, as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (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, http://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. 

The Commission is also offering an 

opportunity for a hearing with respect to 

the issuance of the amendment. 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, 

and electronically on the Internet at the 

NRC Web site, http://www.nrc.gov/ 

reading-rm/doc-collections/cfr/. If there 

are problems in accessing the document, 

contact the PDR Reference staff at 1 

(800) 397–4209, (301) 415–4737, or by 

e-mail to pdr@nrc.gov. If a request for a 

hearing or petition for leave to intervene 

is filed by the above date, 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 identify the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

72680 Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 

1To the extent that the applications contain 

attachments and supporting documents that are not 

publicly available because they are asserted to 

contain safeguards or proprietary information, 

petitioners desiring access to this information 

should contact the applicant or applicant’s counsel 

and discuss the need for a protective order. 

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 intends to 

rely in proving the contention at the 

hearing. The petitioner must also 

provide references to those specific 

sources and documents of which the 

petitioner is aware and on which the 

petitioner intends to rely to establish 

those facts or expert opinion. The 

petition must include sufficient 

information to show that a genuine 

dispute exists with the applicant on a 

material issue of law or fact.1

Contentions shall be limited to matters 

within the scope of the amendment 

under consideration. The contention 

must be one which, if proven, would 

entitle the petitioner to relief. A 

petitioner/requestor who fails to satisfy 

these requirements with respect to at 

least one contention will not be 

permitted to participate as a party. 

Each contention shall be given a 

separate numeric or alpha designation 

within one of the following groups: 

1. Technical—primarily concerns/ 

issues relating to technical and/or 

health and safety matters discussed or 

referenced in the applications. 

2. Environmental—primarily 

concerns/issues relating to matters 

discussed or referenced in the 

environmental analysis for the 

applications. 

3. Miscellaneous—does not fall into 

one of the categories outlined above. 

As specified in 10 CFR 2.309, if two 

or more petitioners/requestors seek to 

co-sponsor a contention, the petitioners/ 

requestors shall jointly designate a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. If a petitioner/requestor 

seeks to adopt the contention of another 

sponsoring petitioner/requestor, the 

petitioner/requestor who seeks to adopt 

the contention must either agree that the 

sponsoring petitioner/requestor shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring petitioner/requestor a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. 

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. Since the Commission has 

made a final determination that the 

amendment involves no significant 

hazards consideration, if a hearing is 

requested, it will not stay the 

effectiveness of the amendment. Any 

hearing held would take place while the 

amendment is in effect. 

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 

email 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 or 

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). 

PPL Susquehanna, LLC, Docket No. 50–

387, Susquehanna Steam Electric 

Station, Unit 1 (SSES–1), Luzerne 

County, Pennsylvania 

Date of amendment request: October 

14, 2005, as supplemented on October 

21 and November 2, 2005. 

Description of amendment request: 

The amendment changed the SSES–1 

Technical Specifications (TSs) by 

revising the SSES–1 Cycle 14 Minimum 

Critical Power Ratio Safety Limit in TS 

Section 2.1.1.2 from 1.08 to 1.09. 

Date of issuance: November 10, 2005. 

Effective date: November 10, 2005. 

Amendment No.: 227. 

Facility Operating License No. NPF–

14: Amendment revised the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. October 24, 

2005 (70 FR 61475). The notice 

provided an opportunity to submit 

comments on the Commission’s 

proposed NSHC determination. No 

comments have been received. The 

notice also provided an opportunity to 

request a hearing by December 22, 2005, 

but indicated that if the Commission 

makes a final NSHC determination, any 

such hearing would take place after 

issuance of the amendment. The 

Commission’s related evaluation of the 

amendment, finding of exigent 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated November 

10th 2005. The supplemental letters 

dated October 21 and November 2, 

2005, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the Nuclear Regulatory Commission 

staff’s original proposed no significant 

hazards consideration determination. 

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. 

Virginia Electric and Power Company, 

Docket No. 50–338, North Anna Power 

Station, Unit No. 1 (North Anna 1), 

Louisa County, Virginia 

Date of amendment request: 

November 3, 2005, as supplemented by 

letter dated November 4, 2005. 

Description of amendment request: 

This amendment allows a temporary 7- 

day Completion Time to repair a weld 

leak that was discovered on the lowhead safety injection (LHSI) suction 

pump piping. This change is needed to 

prevent an unnecessary plant transient 

and unscheduled shutdown of North 

Anna 1. 

Date of issuance: November 4, 2005. 

Effective date: As of the date of 

issuance and is applicable until the ‘‘A’’

train of the Unit 1 LHSI system is 

returned to operable status or until 

November 9, 2005, at 0330 hours, 

whichever occurs first. 

Federal Register / Vol. 70, No. 233 / Tuesday, December 6, 2005 / Notices 72681 

1CRUD is an abbreviation of Chalk River 

Unknown Deposit, a generic term for various 

residues deposited on fuel rod surfaces, originally 

coined by Atomic Energy of Canada, Ltd. to 

describe deposits observed on fuel removed from 

the test reactor at Chalk River. 

Amendment No.: 246. 

Renewed Facility Operating License 

No. NPF–4: Amendment revises the 

Technical Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): No. The 

Commission’s related evaluation of the 

amendment, finding of emergency 

circumstances, state consultation, and 

final NSHC determination are contained 

in a Safety Evaluation dated November 

4, 2005. 

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 

Marinos. 

Dated at Rockville, Maryland, this 28th day 

of November, 2005. 

For the Nuclear Regulatory Commission. 

Catherine Haney, Director, 

Division of Operating Reactor Licensing, 

Office of Nuclear Reactor Regulation. 

[FR Doc. 05–23553 Filed 12–5–05; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Notice of Availability of Documents 

Regarding Spent Fuel Transportation 

Package Response to the Baltimore 

Tunnel Fire Scenario 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Notice of availability. 

FOR FURTHER INFORMATION CONTACT: 

Allen Hansen, Thermal Engineer, 

Criticality, Shielding and Heat Transfer 

Section, Spent Fuel Project Office, 

Office of Nuclear Material Safety and 

Safeguards, U.S. Nuclear Regulatory 

Commission, Washington, DC 20005–

0001. Telephone: (301) 415–1390; fax 

number: (301) 415–8555; e-mail: 

agh@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

Under contract with the Nuclear 

Regulatory Commission (NRC), the 

Pacific Northwest National Laboratory 

prepared the draft NUREG/CR–6886 

report, ‘‘Spent Fuel Transportation 

Package Response to the Baltimore 

Tunnel Fire (BTF) Scenario.’’ The BTF 

was chosen for the study because it 

represents a severe historical accident, 

even though it is a very low frequency 

event. This NUREG/CR documents the 

thermal analyses of three different spent 

fuel transportation packages exposed to 

the BTF scenario: Transnuclear’s TN–

68, Holtec’s HI-STAR 100 and the 

NAC’s LWT. 

To date comments have been received 

from the State of Nevada, Office of the 

Governor, Agency For Nuclear Projects 

and the Western Interstate Energy 

Board. These comments do not need to 

be re-submitted. 

The format of this NUREG/CR has 

been modified since original posting on 

the NRC Electronic Reading Room at 

http://www.nrc.gov/reading-rm/ 

adams.html in September 2005. The 

modified draft NUREG/CR is now 

posted on the NRC Web site at the 

following URLs: 

http://www.nrc.gov/reading-rm/doccollections/nuregs/ 

docs4comment.html. 

http://www.nrc.gov/reading-rm/doccollections/nuregs/contract/cr6886/. 

These links include access to the formal 

comment template. 

The results of this study strongly 

indicate that neither spent nuclear fuel 

(SNF) particles nor fission products 

would be released from a spent fuel 

shipping cask involved in a severe 

tunnel fire such as the Baltimore Tunnel 

Fire. None of the three cask designs 

analyzed for the Baltimore Tunnel fire 

scenario experienced internal 

temperatures that would result in 

rupture of the fuel cladding. Therefore, 

the radioactive material (i.e., SNF 

particles or fission products) would be 

retained within the fuel rods. 

For two of the casks, the TN–68 and 

the NAC–LWT, the maximum 

temperatures experienced in the regions 

of the lid, vent and drain ports exceeded 

the seals’ rated service temperatures, 

making it possible to get a small release 

from the CRUD 1 that might spall off of 

the surfaces of the fuel rods. However, 

any release is expected to be very small 

due to a number of factors. These 

include: (1) The tight clearances 

maintained between the lid and cask 

body; (2) the low pressure differential 

between the cask interior and the 

outside; (3) the tendency of the small 

clearances to plug; and (4) the tendency 

of CRUD particles to settle or plate out. 

The potential releases calculated in 

Chapter 8 for the TN–68 rail cask and 

the NAC–LWT truck cask indicate that 

the release of CRUD from either cask, if 

any, would be very small. There would 

be no release from the HI–STAR 100 

because the inner welded canister 

remains leak tight. 

II. Summary 

The purpose of this notice is to 

provide the public an opportunity to 

review and comment on the Draft 

NUREG/CR–6886 thermal analyses, the 

consequence analyses and the 

conclusions. 

III. Further Information 

The draft NUREG/CR can also be 

viewed at the NRC’s Electronic Reading 

Room at http://www.nrc.gov/readingrm/adams.html. From this site you can 

access the NRC’s Agencywide 

Document Access and Management 

System (ADAMS), which provides text 

and image files of NRC’s public 

documents. The ADAMS accession 

number for the edited (format only) 

NUREG is ML053200024. This file is in 

‘‘black and white.’’ The original draft is 

in color and can be accessed at the 

following accession numbers: 

NUREG/CR Files 

ADAMS 

accession 

No. 

Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario .................................................................... ML052500391 

Appendix A—Material Properties for COBRA–SFS Model of TN–68 Package ................................................................................. ML052490246 

Appendix B—Material Properties for ANSYS Model of HI–STAR 100 Package ............................................................................... ML052490258 

Appendix C—Material Properties for ANSYS Model of Legal Weight Truck Package ...................................................................... ML052490264 

Appendix D—Blackbody View Factors for COBRA–SFS Model of TN–68 Package ......................................................................... ML052490268 

Appendix E—HOLTEC HI–STAR 100 Component Temperature Distributions .................................................................................. ML052490270