67744 Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 

Environment Assessment (Closed—

Ex. 1) 

* 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. 

* * * * * 

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: November 3, 2005. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 05–22316 Filed 11–4–05; 11:02 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 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 October 14, 

2005 to October 27, 2005. The last 

biweekly notice was published on 

October 25, 2005 (70 FR 61655). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. Within 60 days after the 

date of publication of this notice, the 

licensee may file a request for a hearing 

with respect to issuance of the 

amendment to the subject facility 

operating license and any person whose 

interest may be affected by this 

proceeding and who wishes to 

participate as a party in the proceeding 

must file a written request for a hearing 

and a petition for leave to intervene. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules and 

Directives Branch, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

a.m. to 4:15 p.m. Federal workdays. 

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area 01F21, 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 

Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 67745 

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–

4209, (301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1 (HNP), 

Wake and Chatham Counties, North 

Carolina 

Date of amendment request: August 

18, 2005. 

Description of amendment request: 

The amendment will allow the use of 

fire-resistive electrical cable, which has 

been demonstrated to provide an 

equivalent level of protection as would 

be provided by 3-hour and 1-hour rated 

electrical cable raceway fire barriers, for 

the protection of safe shutdown 

electrical cable. 

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. 

Operation of HNP in accordance with the 

proposed amendment does not increase the 

probability or consequences of accidents 

previously evaluated. The Final Safety 

Analysis Report (FSAR) documents the 

analyses of design basis accidents (DBA) at 

HNP. Any scenario or previously analyzed 

accidents that result in offsite dose were 

evaluated as part of this analysis. The 

proposed amendment does not adversely 

affect accident initiators nor alter design 

assumptions, conditions, or configurations of 

the facility. The proposed amendment does 

not adversely affect the ability of structures, 

systems, or components (SSCs) to perform 

their design function. SSCs required to safely 

shut down the reactor and to maintain it in 

a safe shutdown condition remain capable of 

performing their design functions. 

The purpose of this amendment is to 

assure that redundant trains of safe shutdown 

(SSD) control circuits remain protected from 

damage in the event of a postulated fire. The 

proposed amendment revises the Final Safety 

Analysis Report (FSAR) to use three-hour 

fire-resistive electrical cable, which has been 

demonstrated to provide an equivalent level 

of protection as would be provided by threehour and one-hour rated electrical cable 

raceway fire barriers, for the protection of 

67746 Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 

SSD electrical cables. Based on the above, 

SSD control circuit protection is maintained 

by this amendment. 

Therefore, this amendment 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. 

Operation of HNP in accordance with the 

proposed amendment does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. The FSAR documents the analyses 

of design basis accidents (DBA) at HNP. Any 

scenario or previously analyzed accidents 

that result in offsite dose were evaluated as 

part of this analysis. The proposed 

amendment does not change or affect any 

accident previously evaluated in the FSAR, 

and no new or different scenarios are created 

by the proposed amendment. The proposed 

amendment does not adversely affect 

accident initiators nor alter design 

assumptions, conditions, or configurations of 

the facility. The proposed amendment does 

not adversely affect the ability of SSCs to 

perform their design function. SSCs required 

to safely shut down the reactor and to 

maintain it in a safe shutdown condition 

remain capable of performing their design 

functions. 

The purpose of this amendment is to 

assure that redundant trains of Safe 

Shutdown (SSD) control circuits remain 

protected from damage in the event of a 

postulated fire. The proposed amendment 

revises the Final Safety Analysis Report 

(FSAR) to use three-hour fire-resistive 

electrical cable, which has been 

demonstrated to provide an equivalent level 

of protection as would be provided by threehour and one-hour rated electrical cable 

raceway fire barriers, for the protection of 

SSD electrical cables. Based on the above, 

SSD control circuit protection is maintained 

by this amendment. 

Therefore, this amendment 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 of HNP in accordance with the 

proposed amendment does not involve a 

significant reduction in a margin of safety. 

The proposed amendment does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed amendment does not 

adversely affect existing plant safety margins 

or the reliability of equipment assumed to 

mitigate accidents in the FSAR. The 

proposed amendment does not adversely 

affect the ability of SSCs to perform their 

design function. SSCs required to safely shut 

down the reactor and to maintain it in a safe 

shutdown condition remain capable of 

performing their design functions. 

The purpose of this amendment is to 

assure that redundant trains of Safe 

Shutdown (SSD) control circuits remain 

protected from damage in the event of a 

postulated fire. The proposed amendment 

revises the Final Safety Analysis Report 

(FSAR) to use three-hour fire-resistive 

electrical cable, which has been 

demonstrated to provide an equivalent level 

of protection as would be provided by threehour and one-hour rated electrical cable 

raceway fire barriers, for the protection of 

SSD electrical cables. Based on the above, 

SSD control circuit protection is maintained 

by this amendment. 

Therefore, this 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: David T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1, Wake and 

Chatham Counties, North Carolina 

Date of amendment request: 

September 1, 2005. 

Description of amendment request: 

The amendment will add Technical 

Specification (TS) 3.7.14, ‘‘Fuel Storage 

Pool Boron Concentration’’ and revise 

TS 5.6, ‘‘Fuel Storage.’’ The proposed 

changes are related to requirements for 

ensuring adequate subcriticality margin 

in the spent fuel storage pools. TS 5.6.1 

is being revised to include the design 

requirements for dry storage of new fuel. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes do not modify the 

facility. The accident previously analyzed for 

the spent fuel pool is a fuel handling 

accident. The proposed change applies 

administrative controls for maintaining the 

required boron concentration in the spent 

fuel storage pools, revises acceptance criteria 

and storage arrangements for fuel storage in 

PWR [pressurized-water reactor] ‘‘flux trap’’

style racks and adds acceptance criteria for 

dry storage of new fuel to the Technical 

Specifications. The controls on spent fuel 

pool boron and dry storage of new fuel have 

previously been implemented but are being 

added to the Technical Specifications as 

requirements. The proposed change applies 

new acceptance criteria for criticality safety 

of fuel storage in PWR ‘‘flux trap’’ style racks 

in Pools ‘‘A’’ and ‘‘B.’’ The new acceptance 

criteria require new administrative controls 

on the placement of fuel in Pools ‘‘A’’ and 

‘‘B.’’ Similar administrative controls have 

previously been placed on fuel stored in 

Pools C and D. These changes will eliminate 

the dependence on Boraflex in the PWR ‘‘flux 

trap’’ style storage racks. These changes do 

not impact the probability of having a fuel 

handling accident and do not impact the 

consequences of a fuel handling accident. 

Therefore, this amendment 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. 

No change is being made to the acceptance 

criteria of the dry storage of new fuel. These 

criteria are being added to Technical 

Specification Section 5.6.1. Detailed analyses 

have been performed to ensure a criticality 

accident in Pools ‘‘A’’ and ‘‘B’’ is not a 

credible event. The events that could lead to 

a criticality accident are not new. These 

events include a fuel mis-positioning event, 

a fuel drop event, and a boron dilution event. 

The proposed changes do not impact the 

probability of any of these events. The 

detailed criticality analyses performed 

demonstrate that criticality would not occur 

following any of these events. For the more 

likely event, such as a fuel mis-positioning 

event, the acceptance criteria for keff remains 

less than or equal to 0.95. For the unlikely 

event that the spent fuel storage pool boron 

concentration was reduced to zero, keff

remains less than 1.0. 

Therefore, a criticality accident remains 

‘‘not credible,’’ and this amendment 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. 

Incorporation of acceptance criteria for dry 

storage of new fuel into TS 5.6.1 does not 

involve a reduction in the margin of safety. 

The new fuel storage condition continues to 

meet keff ≤ 0.95 during normal conditions and 

keff ≤ 0.98 under optimal moderation 

conditions. 

The proposed changes for storage of new 

and irradiated fuel in Pools ‘‘A’’ and ‘‘B’’

continue to provide the controls necessary to 

ensure a criticality event could not occur in 

the spent fuel storage spool. The acceptance 

criteria are consistent with the acceptance 

criteria specified in 10 CFR 50.68, which 

provide an acceptable margin of safety with 

regard to the potential for a criticality event. 

Therefore, this amendment does not 

involve a significant reduction in a margin of 

safety. 

Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 67747 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: 

September 26, 2005. 

Description of amendment request: 

The proposed amendment will revise 

the analysis method used for the largebreak loss-of-coolant accident 

(LBLOCA) by incorporating the use of a 

new approach (ASTRUM) for the 

treatment of parameter uncertainties. 

The new approach is described in 

Westinghouse Topical Report WCAP–

16009–P–A, approved by the NRC on 

November 5, 2004. 

Changes to the Technical 

Specifications to reflect the proposed 

use of ASTRUM in LBLOCA analysis 

consist of revisions to the list of 

references provided in Technical 

Specification Section 5.6.5, Core 

Operating Limits Report. 

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 modifies the analysis 

methodology used to account for the 

variation in parameters that are used for the 

safety analysis of the LBLOCA. This 

proposed change has no effect on the design 

or operation of plant equipment. Use of the 

new methodology will revise the results of 

the current analysis, but there will be no 

change in initiating events for this accident 

scenario or the ability of the plant equipment 

or plant operators to respond. 

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 

modifications to existing plant equipment or 

the installation of any new equipment. The 

proposed change only affects the analysis 

methodology that is used to evaluate the 

response of existing plant equipment to the 

LBLOCA scenario. Plant operating and 

emergency procedures that are in place for 

the LBLOCA scenario are also not being 

changed by this proposed amendment. This 

proposed change does not create new failure 

modes or malfunctions of plant equipment 

nor is there a new credible failure 

mechanism. 

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 license amendment revises 

the analysis methodology which is used to 

assess the impact of the LBLOCA scenario 

with respect to established acceptance 

criteria. Margins of safety for LBLOCA 

include quantitative limits for fuel 

performance established in 10 CFR 50.46. 

These acceptance criteria and the associated 

margins of safety are not being changed. The 

evaluation of the LBLOCA scenario, using the 

proposed new methodology must still meet 

the existing established acceptance criteria. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Section Chief: Richard J. Laufer. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Date of amendment request: April 4, 

2005. 

Description of amendment request: 

The proposed amendments would 

revise the maximum and minimum 

allowable values for the degraded 

voltage function of the 4160 volt 

essential service system (ESS) bus 

under-voltage instrumentation. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise the Technical 

Specifications (TS) maximum and minimum 

allowable values for the degraded voltage 

protection function and implement the use of 

automatic load tap changers (LTCs) on 

transformers that provide power to safetyrelated equipment. The only accident 

previously evaluated for which the 

probability is potentially affected by these 

changes is the loss of offsite power (LOOP). 

An allowable value for the degraded voltage 

protection function that is too high could 

cause the emergency buses to transfer to the 

emergency diesel generators (EDG) and thus 

increase the probability of a LOOP. The 

allowable value for the degraded voltage 

protection function has been revised in 

accordance with an NRC-approved setpoint 

methodology and will continue to ensure that 

the degraded voltage protection function 

actuates when required, but does not actuate 

prematurely to cause a LOOP. 

A failure of an LTC while in automatic 

operation mode that results in decreased 

voltage to the ESS buses could also cause a 

LOOP. This could occur in two ways. A 

failure of the LTC controller that results in 

rapidly decreasing the voltage to the 

emergency buses is the most severe failure 

mode. However, a backup controller is 

provided with the LTC that makes this failure 

highly unlikely. A failure of the LTC 

controller to respond to decreasing grid 

voltage is less severe, since grid voltage 

changes occur slowly. In both of the above 

potential failure modes, operators will take 

manual control of the LTC to mitigate the 

effects of the failure. Thus, the probability of 

a LOOP is not significantly increased. 

The proposed changes will have no effect 

on the consequences of a LOOP, since the 

EDGs provide power to safety related 

equipment following a LOOP. The EDGs are 

not affected by the proposed changes. 

The probability of other accidents 

previously evaluated is not affected, since the 

proposed changes do not affect the way plant 

equipment is operated and thus do not 

contribute to the initiation of any of the 

previously evaluated accidents. The only 

way in which the consequences of other 

previously evaluated accidents could be 

affected is if a failure of the LTC while in 

automatic operation mode caused a sustained 

high voltage which resulted in damage to 

safety related equipment that is used to 

mitigate an accident. Damage due to overvoltage is time-dependent. Since the LTC is 

equipped with a backup controller, and since 

operator action is available to prevent a 

sustained high voltage condition from 

occurring, damage to safety related 

equipment is extremely unlikely, and thus 

the consequences of these accidents are not 

significantly increased. 

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 

67748 Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes involve functions 

that provide offsite power to safety related 

equipment for accident mitigation. Thus, the 

proposed changes potentially affect the 

consequences of previously evaluated 

accidents (as addressed in Question 1), but 

do not result in any new mechanisms that 

could initiate damage to the reactor and its 

principal safety barriers (i.e., fuel cladding, 

reactor coolant system, or primary 

containment). 

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 changes do not affect the 

inputs or assumptions of any of the analyses 

that demonstrate the integrity of the fuel 

cladding, reactor coolant system, or 

containment during accident conditions. The 

allowable values for the degraded voltage 

protection function have been revised in 

accordance with an NRC-approved setpoint 

methodology and will continue to ensure that 

the degraded voltage protection function 

actuates when required, but does not actuate 

prematurely to cause a LOOP. Automatic 

operation of the LTC increases margin by 

reducing the potential for transferring to the 

EDGs during an event. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Thomas S. 

O’Neill, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Gene Y. Suh. 

FPL Energy Seabrook, LLC, Docket No. 

50–443, Seabrook Station, Unit No. 1, 

Rockingham County, New Hampshire 

Date of amendment request: 

September 22, 2005. 

Description of amendment request: 

The proposed amendment would revise 

the Seabrook Station, Unit No. 1 

operating license and Technical 

Specifications to increase the licensed 

rated power level by 1.7 percent from 

3587 megawatts thermal (MWt) to 3648 

MWt. Basis for proposed no significant 

hazards consideration determination: As 

required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The proposed change will not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Seabrook Station performed evaluations of 

the Nuclear Steam Supply System (NSSS) 

and balance of plant systems, components, 

and analyses that could be affected by the 

proposed change. A power uncertainty 

calculation was performed, and the effect of 

increase core thermal power by 1.7 percent 

to 3648 MWt on the Seabrook Station design 

and licensing basis was evaluated. The result 

of the evaluations determined that all 

systems and components continue to be 

capable of performing their design function 

at the MUR [measurement uncertainty 

recapture] core power level of 3648 MWt. An 

evaluation of the accident analyses 

demonstrates that the applicable analyses 

acceptance criteria continue to be met. No 

accident initiators are affected by the MUR 

power uprate and no challenges to any plant 

safety barriers are created by the proposed 

change. 

The proposed change does not affect the 

release paths, the frequency of release, or the 

analyzed source term for any accidents 

previously evaluated in the Seabrook Station 

Updated Final Safety Analysis Report 

(UFSAR). Systems, structures, and 

components required to mitigate transients 

continue to be capable of performing their 

design functions, and thus were found 

acceptable. The reduced uncertainty in the 

feedwater flow input to the power 

calorimetric measurement ensures that 

applicable accident analyses acceptance 

criteria continue to be met, to support 

operation at the MUR core power level of 

3648 MWt. Analyses performed to assess the 

effects of mass and energy remain valid. The 

source term used to assess radiological 

consequences [has] been reviewed and 

determined to bound operation at the MUR 

core power level. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change will not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

No new accident scenarios, failure 

mechanisms, or single failures are introduced 

as a result of the proposed change. The 

installation of the Caldon LEFM CheckPlusTM

System has been analyzed, and failures of the 

system will have no adverse effect on any 

safety-related system or any systems, 

structures, and components required for 

transient mitigation. Systems, structures, and 

components previously required for the 

mitigation of a transient continue to be 

capable of fulfilling their intended design 

functions. The proposed change has no 

adverse affect on any safety-related system or 

component and does not change the 

performance or integrity of any safety-related 

system. 

The proposed change does not adversely 

affect any current system interfaces or create 

any new interfaces that could result in an 

accident or malfunction of a different kind 

than previously evaluated. Operating at a 

core power level of 3648 MWt does not create 

any new accident initiators or precursors. 

The reduced uncertainty in the feedwater 

flow input to the power calorimetric 

measurement ensures that applicable 

accident analyses acceptance criteria 

continue to be met, to support operation at 

the MUR core power level of 3648 MWt. 

Credible malfunctions continue to be 

bounded by the current accident analyses of 

record or evaluations that demonstrate that 

applicable criteria continue to be met. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. The proposed change will not involve a 

significant reduction in a margin [of] safety. 

The margins of safety associated with the 

MUR are those pertaining to core thermal 

power. These include those associated with 

the fuel cladding, Reactor Coolant System 

pressure boundary, and containment barriers. 

An engineering evaluation of the 1.7 percent 

increase in core thermal power from 3587 

MWt to 3648 MWt was performed. The 

current licensing bases analyzed core power 

is 3659 MWt. The analyzed core power level 

of 3659 MWt bounds the NSSS thermal and 

hydraulic parameters at the MUR core power 

level of 3648 MWt. The NSSS systems and 

components were evaluated at the MUR core 

power level and it was determined that the 

NSSS systems and components continue to 

operate satisfactorily at the MUR power level. 

The NSSS accident analyses were evaluated 

at the MUR core power level of 3648 MWt. 

In all cases, the accident analyses at the MUR 

core power level of 3648 MWt were bounded 

by the current licensing bases analyzed core 

power level of 3659 MWt. As such, the 

margins of safety continue to be bounded by 

the current analyses of record for this change. 

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: M. S. Ross, 

Florida Power & Light Company, P.O. 

Box 14000, Juno Beach, FL 33408–0420. 

NRC Section Chief: Darrell J. Roberts. 

FPL Energy Seabrook, LLC, Docket No. 

50–443, Seabrook Station, Unit No. 1, 

Rockingham County, New Hampshire 

Date of amendment request: 

September 29, 2005. 

Description of amendment request: 

The proposed amendment would revise 

the Seabrook Station, Unit No. 1, 

Technical Specifications (TSs) to permit 

a one-time, six-month extension to the 

currently approved 15-year test interval 

for the containment integrated leak rate 

test. 

Basis for proposed no significant 

hazards consideration determination: 

Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 67749 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The proposed change [does] not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The probability or consequences of 

accidents previously evaluated in the UFSAR 

[updated final safety analysis report] are 

unaffected by this proposed change. There is 

no change to any equipment response or 

accident mitigation scenario, and this change 

results in no additional challenges to fission 

product barrier integrity. The proposed 

change does not alter the design, 

configuration, operation, or function of any 

plant system, structure, or component. As a 

result, the outcomes of previously evaluated 

accidents are unaffected. The proposed 

extension to the containment integrated leak 

rate test (ILRT) interval does not involve a 

significant increase in consequences because, 

as discussed in NUREG 1493, Performance 

Based Containment Leak Rate Test Program, 

Type B and C tests identify the vast majority 

(greater than 95 percent) of all potential 

leakage paths. Further, ILRTs identify only a 

few potential leakage paths that cannot be 

identified through Type B and C testing, and 

leaks found by Type A testing have been only 

marginally greater than existing 

requirements. In addition, periodic 

inspections ensure that any significant 

containment degradation will not go 

undetected. Therefore, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change [does] not create 

the possibility of a new or different kind of 

accident from any [accident] previously 

evaluated. 

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 system. The proposed change 

neither installs or removes any plant 

equipment, nor alters the design, physical 

configuration, or mode of operation of any 

plant structure, system, or component. No 

physical changes are being made to the plant, 

so no new accident causal mechanisms are 

being introduced. The proposed change only 

changes the frequency of performing the 

ILRT; however, the test implementation and 

acceptance criteria are unchanged. Therefore, 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. The proposed change [does] not involve 

a significant reduction in a margin of safety. 

The margin of safety associated with the 

acceptance criteria of any accident is 

unchanged. The proposed change will have 

no affect on the availability, operability, or 

performance of the safety-related systems and 

components. The proposed change does not 

alter the design, configuration, operation, or 

function of any plant system, structure, or 

component. The ability of any operable 

structure, system, or component to perform 

its designated safety function is unaffected by 

this change. NUREG 1493 concluded that 

reducing the frequency of ILRTs to 20 years 

resulted in an imperceptible increase in risk. 

Also, inspections of containment, required by 

the ASME code [American Society of 

Mechanical Engineers Boiler and Pressure 

Vessel Code] and the maintenance rule, 

ensure that containment will not degrade in 

a manner that is only detectable by Type A 

(ILRT) testing. Therefore, the margin of safety 

as defined in the TS is not reduced and 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: M. S. Ross, 

Florida Power & Light Company, P.O. 

Box 14000, Juno Beach, FL 33408–0420. 

NRC Section Chief: Darrell J. Roberts. 

FPL Energy Seabrook, LLC, Docket No. 

50–443, Seabrook Station, Unit No. 1, 

Rockingham County, New Hampshire 

Date of amendment request: 

September 29, 2005. 

Description of amendment request: 

The proposed amendment would revise 

the Seabrook Station, Unit No. 1 

Technical Specifications to permit a 

change in the steam generator tube 

inspection requirements to include a 

sampling of the bulges and overexpansions for portions of the steam 

generator tubes within the hot leg 

tubesheet region. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The previously analyzed accidents are 

initiated by the failure of plant structures, 

systems, or components. The proposed 

changes that alter the steam generator 

inspection criteria do not have a detrimental 

impact on the integrity of any plant structure, 

system, or component that initiates an 

analyzed event. The proposed changes will 

not alter the operation of, or otherwise 

increase the failure probability of any plant 

equipment that initiates an analyzed 

accident. 

Of the applicable accidents previously 

evaluated, the limiting transients with 

consideration to the proposed changes to the 

steam generator tube inspection criteria, are 

the steam generator tube rupture (SGTR) 

event and the steam line break (SLB) 

accident. 

During the SGTR event, the required 

structural integrity margins of the steam 

generator tubes will be maintained by the 

presence of the steam generator tubesheet 

area. Tube rupture in tubes with cracks in the 

tubesheet is precluded by the constraint 

provided by the tubesheet. This constraint 

results from the hydraulic expansion process, 

thermal expansion mismatch between the 

tube and tubesheet and from the differential 

pressure between the primary and secondary 

side. Based on this design, the structural 

margins against burst, as discussed in 

Regulatory Guide (RG) 1.121, ‘‘Bases for 

Plugging Degraded PWR [pressurized-water 

reactor] Steam Generator Tubes,’’ are 

maintained for both normal and postulated 

accident conditions. 

At normal operating pressures, leakage 

from primary water stress corrosion cracking 

(PWSCC) below the proposed limited 

inspection depth is limited by both the tubeto-tubesheet crevice and the limited crack 

opening permitted by the tubesheet 

constraint. Consequently, negligible normal 

operating leakage is expected from cracks 

within the tubesheet region. The 

consequences of an SGTR event are affected 

by the primary-to-secondary leakage flow 

during the event. Primary-to-secondary 

leakage flow through a postulated ruptured 

tube is not affected by the proposed changes 

since the tubesheet enhances the tube 

integrity in the region of the hydraulic 

expansion by precluding tube deformation 

beyond its initial hydraulically-expanded 

outside diameter. 

Furthermore, the proposed changes do not 

affect other systems, structures, components 

or operational features. Therefore, the 

proposed changes result in no significant 

increase in the probability of the occurrence 

of a SGTR accident. 

The probability of a[n] SLB accident is 

unaffected by the potential failure of a steam 

generator tube as this failure is not an 

initiator for a[n] SLB accident. 

The consequences of a[n] SLB accident are 

also not significantly affected by the 

proposed changes. During a[n] SLB accident, 

the reduction in pressure above the tubesheet 

on the shell side of the steam generator 

creates an axially uniformly distributed load 

on the tubesheet due to the reactor coolant 

system pressure on the underside of the 

tubesheet. The resulting bending action 

constrains the tubes in the tubesheet thereby 

restricting primary-to-secondary leakage 

below the midplane. 

Primary-to-secondary leakage from tube 

degradation in the tubesheet area during the 

limiting accident (i.e., a[n] SLB) is limited by 

flow restrictions resulting from the crack and 

tube-to-tubesheet contact pressures that 

provide a restricted leakage path above the 

indications and also limit the degree of 

potential crack face opening as compared to 

free span indications. The primary-tosecondary leak rate during postulated SLB 

accident conditions would be expected to be 

less than that during normal operation for 

indications near the bottom of the tubesheet 

(i.e., including indications in the tube end 

welds). This conclusion is based on the 

67750 Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 

observation that while the driving pressure 

causing leakage increases by approximately a 

factor of (two) 2, the flow resistance 

associated with an increase in tube-totubesheet contact pressure, during a[n] SLB 

accident, increases by approximately a factor 

of 2.5. While such a leakage decrease is 

logically expected, the postulated accident 

leak rate could be conservatively bounded by 

twice the normal operating leak rate even if 

the increase in contact pressure is ignored. 

Since normal operating leakage (spiking) is 

limited to less that 0.104 gpm (150 gpd) for 

continued power operation per station 

operating procedure OS 1227.02, ‘‘Steam 

Generator Tube Leak,’’ the associated 

accident condition leak rate, assuming all 

leakage to be from lower tube sheet 

indications, would be bound by 0.208 gpm 

(twice normal operating leak rate). This value 

is well within the assumed accident leakage 

rate of 0.347 gpm discussed in the Seabrook 

Station Updated Safety Analysis Report, 

Section 15.1.5 ‘‘Steam System Piping 

Failure.’’ Hence it is reasonable to omit any 

consideration of inspection of the tube, tube 

end weld, bulges / overexpansions or other 

anomalies below 17 inches from the top of 

the hot leg tubesheet. Therefore, the 

consequences of a[n] SLB accident remain 

unaffected. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed changes do not create the 

possibility of a new or different kind of 

accident from any [accident] previously 

evaluated. 

The proposed changes do not introduce 

any new equipment, create new failure 

modes for existing equipment, or create any 

new limiting single failures. Plant operation 

will not be altered, and all safety functions 

will continue to perform as previously 

assumed in accident analyses. Therefore, the 

proposed changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. The proposed changes do not involve a 

significant reduction in the margin of safety. 

The proposed changes maintain the 

required structural margins of the steam 

generator tubes for both normal and accident 

conditions. Nuclear Energy Institute (NEI) 

97–06, ‘‘Steam Generator Program 

Guidelines,’’ and NRC Regulatory Guide (RG) 

1.121, ‘‘Bases for Plugging Degraded PWR 

Steam Generator Tubes,’’ are used as the 

bases in the development of the limited hot 

leg tubesheet inspection depth methodology 

for determining that steam generator tube 

integrity considerations are maintained 

within acceptable limits. RG 1.121 describes 

a method acceptable to the NRC for meeting 

General Design Criteria (GDC) 14, ‘‘Reactor 

Coolant Pressure Boundary,’’ GDC 15, 

‘‘Reactor Coolant System Design,’’ GDC 31, 

‘‘Fracture Prevention of Reactor Coolant 

Pressure Boundary,’’ and GDC 32, 

‘‘Inspection of Reactor Coolant Pressure 

Boundary,’’ by reducing the probability and 

consequences of a SGTR. RG 1.121 concludes 

that by determining the limiting safe 

conditions for tube wall degradation the 

probability and consequences of a SGTR are 

reduced. RG 1.121 uses safety factors on 

loads for tube burst that are consistent with 

the requirements of Section III of the 

American Society of Mechanical Engineers 

(ASME) Code. 

For axially oriented cracking located 

within the tubesheet, tube burst is precluded 

due to the presence of the tubesheet. For 

circumferentially oriented cracking, 

Westinghouse evaluation LTR-CDME–05–

170, ‘‘Limited Inspection of the Steam 

Generator Tube Portion Within the Tubesheet 

at Seabrook Generating Station,’’ defines a 

length of degradation-free expanded tubing 

that provides the necessary resistance to tube 

pullout due to the pressure induced forces, 

with applicable safety factors applied. 

Application of the limited hot leg tubesheet 

inspection criteria will preclude 

unacceptable primary-to-secondary leakage 

during all plant conditions. The methodology 

for determining leakage provides for large 

margins between calculated and actual 

leakage values in the proposed limited hot 

leg tubesheet inspection depth criteria. 

Therefore, the proposed changes do not 

involve a significant reduction in any margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: M. S. Ross, 

Florida Power & Light Company, P.O. 

Box 14000, Juno Beach, FL 33408–0420. 

NRC Section Chief: Darrell J. Roberts. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendment request: 

September 27, 2005. 

Description of amendment request: 

The amendments proposed by Southern 

Nuclear Operating Company would 

revise the Technical Specifications (TS) 

to eliminate the Power Range Neutron 

Flux-High Negative Rate Reactor Trip 

function, based on the approved 

methodology contained in 

Westinghouse Topical Report WCAP–

11394–P–A, ‘‘Methodology for the 

Analysis of the Dropped Rod Event.’’

The changes will allow the elimination 

of a trip circuitry that is not credited in 

the Farley Nuclear Plant safety analysis, 

and which can result in an unnecessary 

reactor trip. These changes will be 

implemented sequentially, concurrent 

with each unit’s refueling outage during 

which the design change is 

implemented. Additionally, this 

amendment request deletes TS Bases 

text associated with an unconservative 

local Departure from Nucleate Boiling 

Ratio. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed changes do not 

significantly increase the probability or 

consequences of an accident previously 

evaluated in the Updated Final Safety 

Analysis Report (UFSAR). All of the safety 

analyses have been evaluated for impact due 

to this change. The elimination of the Power 

Range Neutron Flux-High Negative Rate 

Reactor Trip function and the elimination of 

text in the TS [Technical Specifications] 

Bases for LC0 3.3.1, page B 3.3.1–1 1, 

associated with an unconservative local 

DNBR [departure from nucleate boiling ratio], 

does not affect the dropped RCCA [Rod 

Cluster Control Assembly] analyses nor any 

other analyses, since it is not credited in any 

of the safety analyses; therefore, the 

probability of an accident has not been 

increased. All dose consequences have been 

evaluated with respect to the proposed 

changes, there is no impact due to the 

proposed change, and all acceptance criteria 

continue to be met. Therefore, these changes 

do not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

No. The proposed changes do not create 

the possibility of a new or different kind of 

accident from any accident already evaluated 

in the UFSAR. No new accident scenarios, 

failure mechanisms or limiting single failures 

are introduced as result of the proposed 

changes. The changes have no adverse effects 

on any safety-related system. Therefore, all 

accident analyses criteria continue to be met 

and these changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

No. The proposed changes do not involve 

a significant reduction in a margin of safety. 

The dropped RCCA(s) event does not credit 

the Power Range Neutron Flux-High Negative 

Rate Reactor Trip function. The conclusion 

presented in the UFSAR Section 15.2.3.3 that 

the DNBR design basis is met for a dropped 

RCCA(s) event remains valid for the 

proposed changes, which are based on the 

NRC approved methodology contained in 

CAP–11394–PA. Additionally, WCAP–

11394–P–A indicates that the analysis for a 

dropped rod event envelops a multiple rod 

drop accident at high power levels, and that 

such an accident will not result in an 

unconservative local DNBR. All applicable 

acceptance criteria continue to be met. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 67751 

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: M. Stanford 

Blanton, Esq., Balch and Bingham, Post 

Office Box 306, 1710 Sixth Avenue 

North, Birmingham, Alabama 35201. 

NRC Section Chief: Evangelos C. 

Marinos. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 

1 and 2, Houston County, Alabama 

Date of amendment request: October 

6, 2005. 

Description of amendment request: 

The amendments proposed by Southern 

Nuclear Operating Company (SNC) 

would revise the Technical 

Specifications (TS) to support a revision 

to the Best Estimate Loss of Coolant 

Accident (BELOCA) for Farley Nuclear 

Plant (FNP). The NRC recently approved 

a new Westinghouse BELOCA 

methodology, Automated Statistical 

Treatment of Uncertainty Method 

(ASTRUM). ASTRUM was submitted in 

WCAP–16009–P. The NRC issued a 

Safety Evaluation Report in a letter 

dated November 5, 2004. Westinghouse 

issued WCAP–16009–P–A in January 

2005. SNC has completed the analysis 

for FNP and the enclosed proposed 

amendment is to incorporate a reference 

to WCAP–16009–P–A in TS section 

5.6.5 Core Operating Limits Report 

(COLR). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No physical plant changes are being made 

as a result of using the Westinghouse Best 

Estimate Large Break LOCA [Loss of Coolant 

Accident] (BELOCA) analysis methodology. 

The proposed TS changes simply involve 

updating the references in TS 5.6.5.b, Core 

Operating Limits Report (COLR), to reference 

the Westinghouse BELOCA analysis 

methodology. The plant conditions assumed 

in the analysis are bounded by the design 

conditions for all equipment in the plant; 

therefore, there will be no increase in the 

probability of a LOCA. The consequences of 

a LOCA are not being increased, since the 

analysis has shown that the Emergency Core 

Cooling System (ECCS) is designed such that 

its calculated cooling performance conforms 

to the criteria contained in 10 CFR 50.46, 

‘‘Acceptance criteria for emergency core 

cooling systems for light-water nuclear power 

reactors.’’ No other accident consequence is 

potentially affected by this change. 

All systems will continue to be operated in 

accordance with current design requirements 

under the new analysis, therefore no new 

components or system interactions have been 

identified that could lead to an increase in 

the probability of any accident previously 

evaluated in the Updated Final Safety 

Analysis Report (UFSAR). No changes were 

required to the Reactor Protection System 

(RPS) or Engineering Safety Features (ESF) 

setpoints because of the new analysis 

methodology. 

Therefore, it is concluded that this change 

does not significantly increase 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? 

There are no physical changes being made 

to the plant as a result of using the 

Westinghouse Best Estimate Large Break 

LOCA analysis methodology. No new modes 

of plant operation are being introduced. The 

configuration, operation and accident 

response of the structures or components are 

unchanged by utilization of the new analysis 

methodology. Analyses of transient events 

have confirmed that no transient event 

results in a new sequence of events that 

could lead to a new accident scenario. The 

parameters assumed in the analysis are 

within the design limits of existing plant 

equipment. 

In addition, employing the Westinghouse 

Best Estimate Large Break LOCA analysis 

methodology does not create any new failure 

modes that could lead to a different kind of 

accident. The design of all systems remains 

unchanged and no new equipment or 

systems have been installed which could 

potentially introduce new failure modes or 

accident sequences. No changes have been 

made to any RPS or ESF actuation setpoints. 

Based on this review, it is concluded that 

no new accident scenarios, failure 

mechanisms or limiting single failures are 

introduced as a result of the proposed 

changes. 

Therefore, the proposed TS changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

It has been shown that the analytic 

technique used in the Westinghouse Best 

Estimate Large Break LOCA analysis 

methodology realistically describes the 

expected behavior of the reactor system 

during a postulated LOCA. Uncertainties 

have been accounted for as required by 10 

CFR 50.46. A sufficient number of LOCAs 

with different break sizes, different locations, 

and other variations in properties have been 

considered to provide assurance that the 

most severe postulated LOCAs have been 

evaluated. The analysis has demonstrated 

that all acceptance criteria contained in 10 

CFR 50.46 paragraph b continue to be 

satisfied. 

Therefore, it is concluded that this change 

does not involve a significant reduction in 

the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: M. Stanford 

Blanton, Esq., Balch and Bingham, Post 

Office Box 306, 1710 Sixth Avenue 

North, Birmingham, Alabama 35201. 

NRC Section Chief: Evangelos C. 

Marino. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 1 

and 2, Burke County, Georgia 

Date of amendment request: January 

27, 2005. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specifications Limiting 

Conditions for Operations 3.3.1, 3.3.2, 

3.3.6, and 3.3.8, by extending the 

Surveillance Test Intervals for the 

Reactor Protection System. 

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? 

The proposed changes to the Completion 

Time, bypass test time, and Surveillance 

Frequencies reduce the potential for 

inadvertent reactor trips and spurious 

actuations and, therefore, do not increase the 

probability of any accident previously 

evaluated. The proposed changes to the 

allowed Completion Time, bypass test time, 

and Surveillance Frequencies do not change 

the response of the plant to any accidents 

and have an insignificant impact on the 

reliability of the reactor trip system and 

engineered safety feature actuation system 

(RTS and ESFAS) signals. The RTS and 

ESFAS will remain highly reliable, and the 

proposed changes will not result in a 

significant increase in the risk of plant 

operation. This is demonstrated by showing 

that the impact on plant safety as measured 

by core damage frequency (CDF) is less than 

1.01E–06 per year and the impact on large 

early release frequency (LERF) is less than 

1.0E–07 per year. In addition, for the 

Completion Time change, the incremental 

conditional core damage probabilities 

(ICCDP) and incremental conditional large 

early release probabilities (ICLERP) are less 

than 5.0E–08. These changes meet the 

67752 Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 

acceptance criteria in Regulatory Guides 

1.174 and 1.177. Therefore, since the RTS 

and ESFAS will continue to perform their 

functions with high reliability as originally 

assumed, and the increase in risk as 

measured by CDF, LERF, ICCDP, and ICLERP 

is within the acceptance criteria of existing 

regulatory guidance, there will not be a 

significant increase in the consequences of 

any accidents. The proposed changes do not 

adversely affect accident initiators or 

precursors nor alter the design assumptions, 

conditions, or configuration of the facility or 

the manner in which the plant is operated 

and maintained. The proposed changes do 

not alter or prevent the ability of structures, 

systems, and components (SSCs) from 

performing their intended function to 

mitigate the consequences of an initiating 

event within the assumed acceptance limits. 

The proposed changes do not affect the 

source term, containment isolation, or 

radiological release assumptions used in 

evaluating the radiological consequences of 

an accident previously evaluated. Further, 

the proposed changes do not increase the 

types or amounts of radioactive effluent that 

may be released offsite, nor significantly 

increase individual or cumulative 

occupational/public radiation exposures. The 

proposed changes are consistent with the 

safety analysis assumptions and resultant 

consequences. Therefore, it is concluded that 

this change does not increase the probability 

of occurrence of a malfunction of equipment 

important to safety. 

2. Does the Proposed Change Create the 

Possibility of a New or Different Kind of 

Accident from any Previously Evaluated? 

The proposed changes do not result in a 

change in the manner in which the RTS and 

ESFAS provide plant protection. The RTS 

and ESFAS will continue to have the same 

setpoints after the proposed changes are 

implemented. There are no design changes 

associated with the license amendment. The 

changes to Completion Time, bypass test 

time, and Surveillance Frequency do not 

change any existing accident scenarios, nor 

create any new or different accident 

scenarios. The changes do not involve a 

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

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation. In 

addition, the changes do not impose any new 

or different requirements or eliminate any 

existing requirements. The changes do not 

alter assumptions made in the safety 

analysis. The proposed changes are 

consistent with the safety analysis 

assumptions and current plant operating 

practice. Therefore, the possibility of a new 

or different malfunction of safety related 

equipment is not created. 

3. Does the Proposed Change Involve a 

Significant Reduction in the Margin of 

Safety? 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings, or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not impacted by these 

changes. Redundant RTS and ESFAS trains 

are maintained, and diversity with regard to 

the signals that provide reactor trip and 

engineered safety features actuation is also 

maintained. All signals credited as primary 

or secondary and all operator actions 

credited in the accident analyses will remain 

the same. The proposed changes will not 

result in plant operation in a configuration 

outside the design basis. The calculated 

impact on risk is insignificant and meets the 

acceptance criteria contained in Regulatory 

Guides 1.174 and 1.177. Although there was 

no attempt to quantify any positive human 

factors benefit due to increased Completion 

Time, bypass test time, and Surveillance 

Frequencies, it is expected there would be a 

net benefit due to a reduced potential for 

spurious reactor trips and actuations 

associated with testing. Therefore, it is 

concluded that this change does not involve 

a significant reduction in the margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Arthur H. 

Domby, Troutman Sanders, 

NationsBank Plaza, Suite 5200, 600 

Peachtree Street, NE., Atlanta, Georgia 

30308–2216. 

NRC Section Chief: Evangelos C. 

Marinos. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant (SQN), Units 1 and 2, 

Hamilton County, Tennessee 

Date of amendment request: 

September 30, 2005 (TS–05–02). 

Description of amendment request: 

The proposed amendment would revise 

the SQN Technical Specification (TS) 

Section 5.0, ‘‘Design Features,’’ to more 

conform with NUREG–1431 Revision 3, 

‘‘Standard Technical Specifications for 

Westinghouse Plants.’’ The proposed 

change included the elimination of 

exclusion area, low population zone, 

and effluent subsections and associated 

figures referred to in Section 5.1, ‘‘Site’’; 

elimination of Section 5.2, 

‘‘Containment’’; elimination of Section 

5.4, ‘‘Reactor Coolant System,’’ as well 

as Section 5.5, ‘‘Meteorological Tower 

Location,’’ and its figure. Lastly, a 

proposed change to the TS 

‘‘Administrative Control’’ section to 

acquire the component cyclic or 

transient limits currently located in the 

‘‘Design Features’’ section. 

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 removal of information and figures 

featuring the locations of the site exclusion 

area, gaseous and liquid effluent boundaries, 

low population zone, and the meteorological 

tower is administrative in nature. Most, if 

not, all of this information is located in other 

licensee control documents, such as the Final 

Safety Analysis Report (FSAR). Congruently, 

the addition of a site location description 

only adds geographical information to the 

TSs. The relocation and revision of the 

component cyclic or transient limits 

requirement does not alter the requirement to 

track and maintain these limits and thus 

considered administrative. This proposed 

amendment involves no technical changes to 

the existing TSs and does not impact 

initiators of analyzed events. The changes 

also do not impact the assumed mitigation of 

accidents or transient events. 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 

change to plant systems, components, or 

operating practices that could result in a 

change in accident generation potential. The 

proposed changes do not impose any new or 

different requirements or eliminate any 

existing requirements. The proposed changes 

do not alter assumptions made in the safety 

analyses and licensing basis. 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 deletion of information and figures 

featuring the locations of the site exclusion 

area, gaseous and liquid effluent boundaries, 

low population zone, and the meteorological 

tower does not affect operational limits or 

functional capabilities of plant systems, 

structures and components. The addition of 

a site location description adds geographical 

information to the TSs. The relocation and 

revision of the component cyclic or transient 

limits requirements also does not affect 

operational limits or functional capabilities 

of plant systems, structures and components. 

These changes pose no effect on margin of 

safety. The TS identified maximum steel 

containment temperature value is not the 

current limiting design value, which is found 

in the FSAR. Its elimination is considered 

administrative in nature and does not result 

in a change of margin of safety to the 

containment design. 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 

Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 67753 

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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Unit 2, 

Somervell County, Texas 

Date of amendment request: April 27, 

2005, as supplemented by letter dated 

July 20, 2005. 

Brief description of amendments: The 

amendment revises Technical 

Specification (TS) 5.6.5, ‘‘Core 

Operating Limits Report,’’ by adding 

topical report WCAP–13060–P–A, 

‘‘Westinghouse Fuel Assembly 

Reconstitution Evaluation 

Methodology,’’ to the list of approved 

methodologies to be used at Comanche 

Peak Steam Electric Station (CPSES), 

Unit 2. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change is administrative in 

nature and as such does not impact the 

condition or performance of any plant 

structure, system or component. The core 

operating limits are established to support 

Technical Specifications 3.1, 3.2, 3.3, 3.4, 

and 3.9. The core operating limits ensure that 

fuel design limits are not exceeded during 

any conditions of normal operation or in the 

event of any Anticipated Operational 

Occurrence (AOO). The methods used to 

determine the core operating limits for each 

operating cycle are based on methods 

previously found acceptable by the NRC and 

listed in TS section 5.6.5.b. Application of 

these approved methods will continue to 

ensure that acceptable operating limits are 

established to protect the fuel cladding 

integrity during normal operation and AOOs. 

The requested Technical Specification 

change does not involve any plant 

modifications or operational changes that 

could affect system reliability, performance, 

or possibility of operator error. The requested 

change does not affect any postulated 

accident precursors, does not affect any 

accident mitigation systems, and does not 

introduce any new accident initiation 

mechanisms. 

As a result, the proposed change to the 

CPSES Technical Specifications does not 

involve any increase in the probability or the 

consequences of any accident or malfunction 

of equipment important to safety previously 

evaluated since neither accident probabilities 

nor consequences are being affected by this 

proposed administrative change. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change is administrative in 

nature, and therefore does not involve any 

change in station operation or physical 

modifications to the plant. In addition, no 

changes are being made in the methods used 

to respond to plant transients that have been 

previously analyzed. No changes are being 

made to plant parameters within which the 

plant is normally operated or in the 

setpoints, which initiate protective or 

mitigative actions, and no new failure modes 

are being introduced. 

Therefore, the proposed administrative 

change to the CPSES Technical 

Specifications does not create the possibility 

of a new or different kind of accident or 

malfunction of equipment important to safety 

from any accident previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change is administrative in 

nature and does not impact station operation 

or any plant structure, system or component 

that is relied upon for accident mitigation. 

Furthermore, the margin of safety assumed in 

the plant safety analysis is not affected in any 

way by the proposed administrative change. 

Therefore, the proposed change to the 

CPSES Technical Specifications does not 

involve any 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: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: David Terao. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: January 

24, 2005. 

Brief description of amendments: The 

amendments will revise the surveillance 

requirements (SRs) for Technical 

Specification 3.7.5, ‘‘Auxilary Feed 

Water (AFW) System.’’ Specifically, a 

note will be added to SRs 3.7.5.1, 

3.7.5.3, and 3.7.5.4 that states, ‘‘AFW 

train(s) may be considered OPERABLE 

during alignment and operation for 

steam generator level control, if it is 

capable of being manually realigned to 

the AFW mode of operation.’’

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change has no impact on the 

probability of any accident previously 

evaluated. The consequences of the limiting 

transients and accidents (full power 

operation) are unaffected by the proposed 

change. At low power sufficient time is 

available to establish auxiliary feedwater 

injection if needed. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures are introduced as a result of 

these changes. There will be no adverse effect 

or challenges imposed on any safety-related 

system as a result of these changes. There are 

no changes in the method by which any 

safety-related plant system performs its safety 

function. Overall protection system 

performance will remain within the bounds 

of the previously performed accident 

analyses and the protection systems will 

continue to function in a manner consistent 

with the plant design basis. The proposed 

changes do not affect the probability of any 

event initiators. The proposed changes do not 

alter any assumptions or change any 

mitigation actions in the radiological 

consequence evaluations in the Final Safety 

Analysis Report (FSAR). 

Therefore, the proposed change[s] do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not affect the 

acceptance criteria for any analyzed event 

nor is there a change to any Safety Analysis 

Limit (SAL). There will be no effect on the 

manner in which safety limits, limiting safety 

system settings, or limiting conditions for 

operation are determined nor will there be 

any effect on those plant systems necessary 

to assure the accomplishment of protection 

functions. There will be no impact on the 

overpower limit, the Departure from Nucleate 

Boiling Ratio (DNBR) limits, the Heat Flux 

Hot Channel Factor (FQ), the Nuclear 

Enthalpy Rise Hot Channel Factor (F’H), the 

Loss of Coolant Accident Peak Centerline 

Temperature (LOCA PCT), peak local power 

density, or any other margin of safety. The 

67754 Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 

radiological dose consequence acceptance 

criteria listed in the Standard Review Plan 

will continue to be met. Since the limiting 

transients and accidents are unaffected, the 

proposed change[s] do not involve a 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: David Terao. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: August 

10, 2005. 

Brief description of amendments: The 

amendments would revise the Technical 

Specification (TS) 5.5.13, ‘‘Diesel Fuel 

Oil Testing Program,’’ to relocate the 

specific American Society for Testing 

and Materials (ASTM) Standard 

reference from the Administrative 

Controls Section of TS to a licenseecontrolled document. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes relocate the specific 

American Society for Testing and Materials 

(ASTM) Standard references from the 

Administrative Controls of TS to a licenseecontrolled document. Since any change to 

the licensee-controlled document will be 

evaluated pursuant to the requirements of 10 

CFR 50.59, ‘‘Changes, tests and 

experiments,’’ no increase in the probability 

or consequences of an accident previously 

evaluated is involved. 

The proposed changes do not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do not alter or prevent 

the ability of structures, systems, and 

components (SSCs) from performing their 

intended function to mitigate the 

consequences of an initiating event within 

the assumed acceptance limits. The proposed 

changes do not affect the source term, 

containment isolation, or radiological release 

assumptions used in evaluating the 

radiological consequences of an accident 

previously evaluated. Further, the proposed 

changes do not increase individual or 

cumulative occupational or public radiation 

exposure. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

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

or different type of equipment will be 

installed) or change in the methods 

governing normal plant operation. In 

addition, the changes do not alter the 

assumptions made in the analysis and 

licensing basis. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The level of safety of facility operation is 

unaffected by the proposed changes since 

there is no change in the intent of the TS 

requirements of assuring fuel oil is of the 

appropriate quality for emergency DG [diesel 

generator] use. The proposed changes 

provide the flexibility needed to utilize stateof-the-art technology in fuel oil sampling and 

analysis methods. 

Therefore the proposed changes do not 

involve a reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: David Terao. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: August 

22, 2005. 

Brief description of amendments: The 

amendments revise Technical 

Specification (TS) 3.7.10, ‘‘Control 

Room Emergency Filtration/ 

Pressurization System (CREFS) and 

Control Room Envelope (CRE),’’ and 

adds new TS 5.5.20, ‘‘Control Room 

Integrity Program,’’ and TS 5.6.11, 

‘‘Control Room Report.’’ In addition the 

amendments update the Final Safety 

Analysis Report to include new 

methods and assumptions as described 

in Regulatory Guide 1.195 for evaluation 

of radiological consequences. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change addresses the Control 

Room Envelope (CRE), including updated 

surveillances for the Control Room 

Emergency Filtration/Pressurization System 

(CREFS) trains and the CRE, a new TS 5.5.20, 

‘‘Control Room Integrity Program,’’ and a 

new TS 5.6.11, ‘‘Control Room Report.’’

These changes are consistent with the 

guidance in Regulatory Guides 1.196 and 

1.197. New methods and assumptions for 

evaluating radiological consequences for 

design basis accidents are adopted consistent 

with NRC Regulatory Guide 1.195. The 

acceptance limits for the Control Room 

Integrity Program are based on these revised 

radiological dose consequences calculations. 

The proposed changes do not adversely affect 

accident initiators or precursors nor alter the 

configuration of the facility. The proposed 

changes do not alter or prevent the ability of 

structures, systems, and components (SSCs) 

from performing their intended function to 

mitigate the consequences of an initiating 

event to within the Regulatory Guide 1.195 

acceptance limits. This activity is a revision 

to the Technical Specifications and the 

supporting radiological dose consequences 

analyses for the control room ventilation 

system which is a mitigating system designed 

to minimize in-leakage into the control room 

and to filter the control room atmosphere to 

protect the control room operators following 

accidents previously analyzed. An important 

part of the system is the control room 

envelope (CRE). The CRE integrity is not an 

initiator or precursor to any accident 

previously evaluated. Therefore the 

probability of occurrence of any accident 

previously evaluated is not increased. 

Performing tests and implementing programs 

that verify the integrity of the CRE and 

control room habitability ensure mitigation 

features are capable of performing the 

assumed function. 

The revised radiological consequences 

analyses, performed using the assumptions 

and methodologies presented in Regulatory 

Guidance 1.195, do not result in significant 

increases in the radiological dose 

consequences to the general public nor to the 

control room operators. All calculated dose 

consequences are within acceptance limits of 

Regulatory Guide 1.195. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes will not alter the 

requirements of the control room ventilation 

Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 67755 

system or its function during accident 

conditions. No new or different accidents 

result from performing the new revised 

actions and surveillances or programs 

required. The changes do not involve a 

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

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation which 

could create the possibility of a new or 

different kind of accident. The proposed 

changes are consistent with the safety 

analysis assumptions and current plant 

operating practices. Therefore, the proposed 

changes do not create the possibility of a new 

or different kind of accident from any 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by these 

changes. The proposed changes will not 

result in plant operation in a configuration 

outside the design basis for an unacceptable 

period of time without mitigating actions. 

The proposed changes do not affect systems 

that are required to respond to safely shut 

down the plant and to maintain the plant in 

a safe shutdown condition. 

Therefore the proposed change does not 

involve a reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section 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 

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. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of application for amendment: 

May 27, 2005. 

Brief description of amendment: The 

amendment revised the technical 

specification (TS) testing frequency for 

the surveillance requirement (SR) in TS 

3.1.4, ‘‘Control Rod Scram Times.’’

Specifically, the change revised the 

frequency for SR 3.1.4.2, ‘‘Control Rod 

Scram Time Testing,’’ from ‘‘120 days 

cumulative operation in MODE 1’’ to 

‘‘200 days cumulative operation in 

MODE 1.’’

Date of issuance: October 25, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 167. 

Facility Operating License No. NPF–

43: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 19, 2005 (70 FR 41443). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 25, 

2005. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of application for amendment: 

May 31, 2005. 

Brief description of amendment: The 

amendment modifies Technical 

Specification (TS) requirements to adopt 

the provisions of Industry/TS Task 

Force (TSTF) change TSTF–359, 

‘‘Increased Flexibility in Mode 

Restraints.’’

Date of issuance: October 20, 2005. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 284. 

Facility Operating License No. DPR–

59: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: August 16, 2005 (70 FR 

48204). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 20, 

2005. 

No significant hazards consideration 

comments received: No. 

Exelon Generating Company, LLC, 

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2, 

Ogle County, Illinois 

Date of application for amendment: 

December 17, 2004, as supplemented by 

letter dated September 28, 2005. 

Brief description of amendment: The 

amendments revised Appendix B, 

Environmental Protection Plan (nonradiological), of the Byron Station 

Facility Operating Licenses. 

Date of issuance: October 18, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 145. 

Facility Operating License Nos. NPF–

37 and NPF–66: The amendments 

revised the Environmental Protection 

Plan. 

Date of initial notice in Federal 

Register: April 12, 2005 (70 FR 19115). 

The supplement dated September 28, 

2005, provided additional information 

that clarified the application, 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 as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

67756 Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 

Safety Evaluation dated October 18, 

2005. 

No significant hazards consideration 

comments received: No. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: October 

25, 2004, as supplement by letter dated 

August 1, 2005. 

Brief description of amendment: The 

amendment revises the required 

channels per trip system for several 

instrument functions contained in 

Technical Specification Tables 3.3.6.1–

1, ‘‘Primary Containment Isolation 

Instrumentation,’’ 3.3.6.2–1, ‘‘Secondary 

Containment Isolation 

Instrumentation,’’ and 3.3.7.1–1 

‘‘Control Room Emergency Filter System 

Instrumentation.’’

Date of issuance: October 27, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 212. 

Facility Operating License No. DPR–

46: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 4, 2005 (70 FR 402). 

The supplement dated August 1, 

2005, provided additional information 

that clarified the application, 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 as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 27, 

2005. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–272, 

Salem Nuclear Generating Station Unit 

No. 1, Salem County, New Jersey 

Date of application for amendment: 

February 23, 2005, as supplemented by 

letters dated August 2, 2005, and 

September 21, 2005. 

Brief description of amendment: The 

amendments revised Technical 

Specifications (TSs) to implement a new 

steam generator tube surveillance 

program that is consistent with the 

program proposed by the TS Task Force 

(TSTF) in TSTF–449. 

Date of issuance: October 14, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 268. 

Facility Operating License No. DPR–

70: The amendments revised the TSs. 

Date of initial notice in Federal 

Register: May 10, 2005 (70 FR 24655). 

Supplements dated August 2, 2005, and 

September 21, 2005, provided 

additional information that clarified the 

application, 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 as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 14, 

2005. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments: 

March 4, 2005, as supplemented August 

2, 2005. 

Brief description of amendments: 

These amendments extend the 

completion time from 1 hour to 24 

hours for Actions ‘‘a’’ and ‘‘b’’ of Salem 

Nuclear Generating Station, Unit Nos. 1 

and 2 Technical Specification (TS) 

3.5.1, ‘‘Accumulators,’’ which requires 

restoration of an accumulator when it 

has been declared inoperable for reasons 

other than boron concentration in the 

accumulator not being within the 

required range. 

Date of issuance: October 14, 2005. 

Effective date: As of the date of 

issuance and to be implemented within 

60 days. 

Amendment Nos.: 267 and 249. 

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: May 24, 2005 (70 FR 29800). 

The August 2, 2005, supplement 

provided clarifying information only 

and did not change the scope of the 

proposed amendment, and did not 

change the initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 14, 

2005. 

No significant hazards consideration 

comments received: No. 

Sacramento Municipal Utility District, 

Docket No. 50–312, Rancho Seco 

Nuclear Generating Station, Sacramento 

County, California 

Date of application for amendment: 

January 24, 2005. 

Brief description of amendment: The 

amendment removes unnecessary and 

obsolete information from the facility 

operating license. 

Date of issuance: September 21, 2005. 

Effective date: September 21, 2005. 

Amendment No.: 132. 

Facility Operating License No. DPR–

54: The amendment revised the License. 

Date of initial notice in Federal 

Register: March 29, 2005 (70 FR 

15947). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 22, 

2005. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of application for amendments: 

August 12, 2004. 

Brief description of amendments: The 

amendments revised Surveillance 

Requirement (SR) 4.7.8.d.3 of the 

Auxiliary Building Gas Treatment 

System (ABGTS) by deleting vacuum 

relief flow requirements. The change 

removes criteria from the SR that is not 

necessary to verify the operability of the 

ABGTS and eliminates confusion 

regarding the basis for the vacuum relief 

flow requirement. 

Date of issuance: August 18, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days. 

Amendment Nos.: 303 and 293. 

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revised 

the technical specifications. 

Date of initial notice in Federal 

Register: October 12, 2004 (69 FR 

60687). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated August 18, 

2005. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application for amendment: 

October 27, 2004, as supplemented by 

letter dated June 17, 2005. 

Brief description of amendment: The 

amendment (1) deleted Conditions 

2.C.(3), 2.C.(4), 2.C.(6) through 2.C.(14), 

Section 2.F, and Attachments 1 and 2, 

and (2) revised Conditions 2.C.(1) and 

2.C.(5), to the facility operating license, 

to reflect completed requirements. In 

addition, the list of attachments and 

appendices to the operating license was 

revised to reflect the deletion of 

Attachments 1 and 2. The proposed 

Federal Register / Vol. 70, No. 215 / Tuesday, November 8, 2005 / Notices 67757 

changes to Technical Specifications 

Table 5.5.9–2, ‘‘Steam Generator Tube 

Inspection,’’ and Table 5.5.9–3, ‘‘Steam 

Generator Repaired Tube Inspection,’’

were also submitted in the licensee’s 

application dated September 17, 2004 

(ULNRC–05056), for the replacement 

steam generator project and were 

approved in Amendment No. 168, 

which was issued in the NRC letter 

dated September 29, 2005. 

Date of issuance: October 25, 2005. 

Effective date: October 25, 2005, and 

shall be implemented within 90 days of 

the date of issuance. 

Amendment No.: 169. 

Facility Operating License No. NPF–

30: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 7, 2004 (69 FR 

70723). The June 17, 2005, 

supplemental letter provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

no significant hazards consideration 

determination. The Commission’s 

related evaluation of the amendment is 

contained in a Safety Evaluation dated 

October 25, 2005. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 31st day 

of October, 2005. 

For the Nuclear Regulatory Commission. 

Ledyard B. Marsh, 

Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 05–22002 Filed 11–7–05; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Notice of Availability of Interim Staff 

Guidance Documents for Fuel Cycle 

Facilities 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Notice of availability. 

FOR FURTHER INFORMATION CONTACT: 

James Smith, Project manager, 

Technical Support Group, Division of 

Fuel Cycle Safety and Safeguards, Office 

of Nuclear Material Safety and 

Safeguards, U.S. Nuclear Regulatory 

Commission, Washington, DC 20005–

0001. Telephone: (301) 415–6459; fax 

number: (301) 415–5370; e-mail: 

jas4@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The Nuclear Regulatory Commission 

(NRC) continues to prepare and issue 

Interim Staff Guidance (ISG) documents 

for fuel cycle facilities. These ISG 

documents provide clarifying guidance 

to the NRC staff when reviewing 

licensee integrated safety analysis, 

license applications or amendment 

requests or other related licensing 

activities for fuel cycle facilities under 

subpart H of 10 CFR part 70. FCSS–ISG–

08 has been issued and is provided for 

information. 

II. Summary 

The purpose of this notice is to 

provide notice to the public of the 

issuance of FCSS–ISG–08, Revision 0, 

which provides guidance to NRC staff to 

address accident sequences that may 

result from natural phenomena hazards 

relative to license application or 

amendment request under 10 CFR Part 

70, Subpart H. FCSS–ISG–08, Revision 

0, has been approved and issued after a 

general revision based on NRC staff and 

public comments on the initial draft. 

III. Further Information 

The document related to this action is 

available electronically at the NRC’s 

Electronic Reading Room at http:// 

www.nrc.gov/reading-rm/adams.html. 

From this site, you can access the NRC’s 

Agencywide Documents Access and 

Management System (ADAMS), which 

provides text and image files of NRC’s 

public documents. The ADAMS 

ascension number for the document 

related to this notice is provided in the 

following table. If you do not have 

access to ADAMS or if there are 

problems in accessing the document 

located in ADAMS, contact the NRC 

Public Document Room (PDR) Reference 

staff at 1–800–397–4209, 301–415–4737, 

or by e-mail to pdr@nrc.gov. 

Interim staff guidance ADAMS 

Accession No. 

FCSS Interim Staff Guidance-08, Revision 0.ML052650305 

This document may also be viewed 

electronically on the public computers 

located at the NRC’s PDR, O 1 F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, MD 20852. The PDR 

reproduction contractor will copy 

documents for a fee. Comments on these 

documents may be forwarded to James 

Smith, Project Manager, Technical 

Support Group, Division of Fuel Cycle 

Safety and Safeguards, Office of Nuclear 

Material Safety and Safeguards, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20005–0001. 

Comments can also be submitted by 

telephone, fax, or e-mail which are as 

follows: Telephone: (301) 415–6459; fax 

number: (301) 415–5370; e-mail: 

jas4@nrc.gov. 

Dated at Rockville, Maryland this 27th day 

of October 2005. 

For the Nuclear Regulatory Commission. 

Melanie A. Galloway, 

Chief, Technical Support Group, Division of 

Fuel Cycle Safety and Safeguards, Office of 

Nuclear Material Safety and Safeguards. 

Attachment—FCSS Interim Staff 

Guidance-08, Revision 0, Natural 

Phenomena Hazards 

Prepared by Division of Fuel Cycle 

Safety and Safeguards, Office of 

Nuclear Material Safety and Safeguards 

Issue 

Additional guidance is required to 

address accident sequences that may 

result from natural phenomena hazards 

in the context of a license application or 

an amendment request under Title 10 

Code of Federal Regulations (10 CFR) 

part 70, subpart H. 

Introduction 

This Interim Staff Guidance (ISG) 

provides additional guidance for 

reviewing the applicant’s (or licensee’s) 

evaluation of natural phenomena 

hazards up to and including ‘‘highly 

unlikely’’ events for both new and 

existing facilities. 

Discussion 

The performance requirements of 10 

CFR 70.61 for facilities processing 

special nuclear materials require that 

individual accident sequences resulting 

in high consequences to workers and 

the public be ‘‘highly unlikely’’ and that 

sequences resulting in intermediate 

consequences to these receptors be 

‘‘unlikely.’’ Although the threshold 

levels that differentiate high 

consequence events from intermediate 

consequence events are established in 

the regulations, the definitions of 

‘‘highly unlikely’’ and ‘‘unlikely’’ are 

not. Definitions of these terms must be 

described in the integrated safety 

analysis (ISA) summary submitted by 

applicants and licensees according to 10 

CFR 70.65(b)(9) and subjected to staff 

approval. Further description of the 

acceptance criteria for the definitions of 

these terms can be found in Chapter 3 

of NUREG–1520, ‘‘Standard Review 

Plan for the Review of a License 

Application for a Fuel Cycle Facility.’’

The implementation of these 

requirements may vary somewhat due to 

different definitions of likelihood 

proposed by different applicants (or