64984 Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov.

2 p.m. Briefing on Emergency 

Preparedness Program Initiatives 

(Closed—Ex. 1) (Contact: Nader 

Mamish (301) 415–1086). *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: 

Dave Gamberoni, (301) 415–1651.

* * * * *

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 were 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 4, 2004. 

Dave Gamberoni, 

Office of the Secretary.

[FR Doc. 04–25024 Filed 11–5–04; 9:34 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 15, 

2004, through October 28, 2004. The last 

biweekly notice was published on 

October 26, 2004 (69 FR 62467). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. Within 60 days after the 

date of publication of this notice, the 

licensee may file a request for a hearing 

with respect to issuance of the 

amendment to the subject facility 

operating license and any person whose 

interest may be affected by this 

proceeding and who wishes to 

participate as a party in the proceeding 

must file a written request for a hearing 

and a petition for leave to intervene. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60-

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules and 

Directives Branch, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

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

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. (Note: 

Public access to ADAMS has been 

temporarily suspended so that security 

reviews of publicly available documents 

may be performed and potentially 

sensitive information removed. Please 

check the NRC Web site for updates on 

the resumption of ADAMS access.) 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 

Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 64985

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/. (Note:

Public access to ADAMS has been 

temporarily suspended so that security 

reviews of publicly available documents 

may be performed and potentially 

sensitive information removed. Please 

check the NRC Web site for updates on 

the resumption of ADAMS access.) If a 

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/

requestor seeks to have litigated at the 

proceeding.

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/

requestor to relief. A petitioner/

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland, 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) e-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HEARINGDOCKET@NRC.GOV; or (4) 

facsimile transmission addressed to the 

Office of the Secretary, U.S. Nuclear 

Regulatory Commission, Washington, 

DC, Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

transmission to 301–415–3725 or by email to OGCMailCenter@nrc.gov. A copy 

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer of 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(I)–(viii). 

For further details with respect to this 

action, see the application for 

amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

reading-rm/adams.html. (Note: Public 

access to ADAMS has been temporarily 

suspended so that security reviews of 

publicly available documents may be 

performed and potentially sensitive 

information removed. Please check the 

NRC Web site for updates on the 

resumption of ADAMS access.) If you 

do not have access to ADAMS or if there 

are problems in accessing the 

documents located in ADAMS, contact 

the NRC 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 amendment request:

September 16, 2004. 

Description of amendment request:

The proposed amendment would revise 

the scope and the frequency of 

Surveillance Requirement (SR) 3.7.6.1 

for verification of one complete cycle of 

each turbine bypass valve (TBV) every 

92 days. The proposed change to SR 

3.7.6.1 would allow a 5 percent stroke 

rather than a complete (100 percent) 

stroke of each TBV, and would extend 

the surveillance frequency from 92 days 

to 120 days. The complete stroke 

verification currently required by SR 

3.7.6.1 once after each entry into MODE 

4 would be retained and renumbered SR 

3.7.6.2. 

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:

64986 Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed change to Technical 

Specification Surveillance Requirement (SR) 

3.7.6.1 will allow a 5% stroke rather than a 

complete (100%) stroke of each turbine 

bypass valve (TBV), and will extend the 

surveillance frequency from 92 days to 120 

days. The requirement to verify one complete 

cycle of each TBV once after each entry into 

MODE 4 will be retained.

The proposed testing requirements will 

provide a level of assurance, equivalent to 

that which now exists, that the TBVs will 

remain operable throughout the operating 

cycle, and that they will be able to perform 

their intended safety function if called upon 

to do so. Additionally, the reduction in the 

potential for plant transients that can result 

from the current testing requirements, will 

more than offset the small increase (less than 

one half of one percent) in TBV failure 

probability per cycle with the proposed 

testing regime. Thus the proposed changes 

will not significantly increase the probability 

of an accident previously evaluated. 

Fermi 2 is analyzed for the increase in 

reactor pressure transient events with the 

assumption that the Main Turbine Bypass 

System (MTBS) is out-of-service. Feedwater 

Controller Failure Upscale represents the 

most limiting event in this analytical 

category, and provides the basis for the 

Minimum Critical Power Ratio (MCPR) 

operating limits that are applicable when the 

MTBS is out of service. Because the proposed 

testing requirements do not alter the 

assumptions for any of the increase in 

pressure transient events, the radiological 

consequences of an accident previously 

evaluated are not increased. 

Therefore, this proposed amendment will 

not involve a significant increase in the 

probability or the consequences of an 

accident previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change does not significantly 

affect the assumed performance of the TBVs, 

nor does it affect any other plant systems, 

structures, or components. In fact, these 

changes reduce the possibility of secondary 

plant transients and the potential for 

recirculation pump runbacks during the 

performance of this SR while at power. The 

proposed changes do not install any new 

plant equipment, nor is installed plant 

equipment being operated in a new or 

different manner. The proposed changes in 

test frequency and methodology will 

continue to ensure that the TBVs remain 

capable of performing their intended safety 

function. Therefore, this proposed change 

will not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The proposed change will modify the 

scope and the frequency of the quarterly full 

stroke test of the TBVs. The operability 

requirements and functional characteristics 

of the TBVs remain unchanged. The 

proposed change to SR 3.7.6.1 from full 

stroke testing to 5% stroke testing, and from 

92 days to 120 days has been evaluated to 

produce only a minimal increase in the 

failure probability of a TBV during each cycle 

(less than one half of one percent). This 

failure probability increase is outweighed by 

the reduction in the potential for plant 

transients resulting from full stroke testing 

during power operation. Both Alstom’s 

sensitivity study, and actual industry 

experience at Ringhals Units 1 and 2 have 

shown that a partial stroke test will ensure 

that the valves remain mechanically operable 

throughout the operating cycle. The Alstom 

study further shows that a partial stroke test 

at 120 days, rather than at 92 days, will 

ensure that the valves remain mechanically 

operable throughout the operating cycle. 

Additionally, retaining the requirement to 

full stroke test each TBV once after each 

entry into MODE 4 will continue to verify 

that the valves are mechanically operable 

prior to their first use following each startup 

from MODE 4. The TBV response times are 

used in determining the effect on the MCPR. 

The surveillance test that ensures the MTBS 

meets the system’s response time limits (SR 

3.7.6.3) is not affected by these proposed 

changes and will continue to be performed at 

its current 18 month frequency. Therefore, 

this proposed change will 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: Peter 

Marquardt, Legal Department, 688 WCB, 

Detroit Edison Company, 2000 2nd 

Avenue, Detroit, Michigan 48226–1279. 

NRC Section Chief: L. Raghavan. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of amendment request: October 

7, 2004. 

Description of amendment request:

The proposed amendment would revise 

the Safety Limit Minimum Critical 

Power Ratio in Technical Specification 

2.1.1.2 to reflect the results of cyclespecific calculations performed for 

Fermi 2 operating Cycles 10 and 11. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The basis of the Safety Limit Minimum 

Critical Power Ratio (SLMCPR) is to ensure 

no mechanistic fuel damage is calculated to 

occur if the limit is not violated. The new 

CPR value preserves the existing margin to 

transition boiling and probability of fuel 

damage is not increased. The derivation of 

the revised SLMCPR for Fermi 2 for 

incorporation into the Technical 

Specifications, and its use to determine plant 

and cycle-specific thermal limits, have been 

performed using NRC approved methods. 

These plant-specific calculations are 

performed each operating cycle and if 

necessary, will require future changes to 

these values based upon revised core designs. 

The revised SLMCPR values do not change 

the method of operating the plant and have 

no effect on the probability of an accident 

initiating event or transient. 

Therefore, this proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change results only from a 

specific analysis for the Fermi 2 Cycle 10 and 

11 cores. This change does not involve any 

new or different methods for operating the 

facility. No new initiating events or 

transients result from these changes. 

Therefore, this proposed amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The new SLMCPR is calculated using NRC 

approved methods with plant and cyclespecific parameters for the Cycle 10 and 11 

core designs. The SLMCPR value is 

established to ensure that greater than 99.9% 

of all fuel rods in the core will avoid 

transition boiling if the limit is not violated, 

thereby preserving the fuel cladding 

integrity. The operating MCPR limit is set 

appropriately above the safety limit value to 

ensure adequate margin when the cyclespecific transients are evaluated. 

Accordingly, the margin of safety is 

maintained with the revised values. 

Therefore, this proposed amendment 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: Peter 

Marquardt, Legal Department, 688 WCB, 

Detroit Edison Company, 2000 2nd 

Avenue, Detroit, Michigan 48226–1279. 

NRC Section Chief: L. Raghavan. 

Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 64987

Duke Energy Corporation, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of amendment request: August 

18, 2004. 

Description of amendment request:

The proposed amendment would 

correct an inadvertent technical 

specification (TS) change associated 

with TS Amendment 184/166 and 182/

164. Licensing Amendment 182/164 

deleted the safety injection steam line 

pressure-low (SLPL) function and all 

concerned references due to redundant 

safety injection signals. This 

amendment was approved on 

September 22, 1998. As part of the 

conversion to standardized TS (STS), 

Amendment 184/166, all concerned 

references to the SLPL function were 

not correctly deleted from STS 3.3.2. 

Specifically, a reference to the SLPL 

function was not deleted from Footnote 

(c) to STS Table 3.3.2–1 and from the 

Basis of STS 3.3.2 Function 4.d.(1). 

Amendment (184/166) was approved on 

September 30, 1998. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

Criterion 1—Does This LAR Involve a 

Significant Increase in the Probability or 

Consequences of an Accident Previously 

Evaluated? 

No. Approval and implementation of this 

LAR will have no effect on accident 

probabilities or consequences since the 

proposed changes are consistent with those 

previously reviewed and approved by the 

NRC in TS Amendment 182/164. 

Criterion 2—Does This LAR Create the 

Possibility of a New or Different Kind of 

Accident From Any Accident Previously 

Evaluated? 

No. This LAR does not involve any 

physical changes to the plant. Therefore, no 

new accident causal mechanisms will be 

generated. The proposed changes are 

consistent with those previously reviewed 

and approved by the NRC in TS Amendment 

182/164. Consequently, plant accident 

analyses will not be affected by these 

changes. 

Criterion 3—Does This LAR Involve a 

Significant Reduction in a Margin of Safety? 

No. Margin of safety is related to the 

confidence in the ability of the fission 

product barriers to perform their design 

functions during and following accident 

conditions. These barriers include the fuel 

cladding, the reactor coolant system, and the 

containment system. The performance of 

these barriers will not be affected by the 

proposed changes since they are consistent 

with those previously reviewed and 

approved by the NRC in TS Amendment 182/

164. Therefore, the proposed changes in this 

license amendment will not result in a 

significant reduction in the facility’s margin 

of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Lisa F. 

Vaughn, Duke Energy Corporation, 422 

South Church Street, Charlotte, North 

Carolina 28201–1006. 

NRC Section Chief: Mary Jane RossLee, Acting. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request: October 

12, 2004. 

Description of amendment request:

The proposed license amendment 

request would change the Final Safety 

Analysis Report (FSAR) to reflect that 

the reactor core isolation cooling (RCIC) 

system is not required to mitigate the 

consequences of the control rod drop 

accident (CRDA). The FSAR revision 

would clarify that although the RCIC 

system is designed to initiate and inject 

into the reactor pressure vessel (RPV) at 

a low water level (L2), the additional 

RPV inventory is not required to prevent 

the accident or to mitigate the 

consequences of the CRDA.

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

This change clarifies, in various sections of 

the FSAR, that RCIC system operation is not 

required in order to mitigate the 

consequences of the CRDA. The proposed 

change involves no changes to plant systems 

or accident analyses. The accident analysis 

for the CRDA demonstrates that core design, 

the control rod pattern controls, and the 

scram signal from the reactor protection 

system (RPS) effectively prevent damage to 

the fuel rods as a result of the dropped rod. 

Furthermore, based on a prescribed source 

term provided from an assumed damage to 

less than 2% fuel in the core, the resulting 

radiological consequences are not affected by 

RCIC operation or failure to operate. As such, 

the change does not affect initiation of 

analyzed events or assumed mitigation of 

accidents or transients. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

This change clarifies, in various sections of 

the FSAR, that the RCIC system operation is 

not required in order to mitigate the 

consequences of the CRDA. The proposed 

change does not involve a physical alteration 

of the plant, add any new equipment, or 

require any existing equipment to be 

operated in a manner different from the 

present design. Therefore, the proposed 

change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

This change clarifies, in various sections of 

the FSAR, that the RCIC system operation is 

not required in order to mitigate the 

consequences of the CRDA. The change has 

no effect on plant systems, operating 

practices or safety analyses assumptions. For 

these reasons, 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: Thomas C. 

Poindexter, Esq., Winston & Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Robert A. Gramm. 

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request:

September 30, 2004. 

Description of amendment request:

The proposed amendment would 

change the existing steam generator tube 

surveillance program to be consistent 

with that being proposed by the 

Technical Specifications Task Force 

(TSTF) in TSTF–449, Draft Revision 2. 

These proposed changes would revise 

the Technical Specifications and Bases 

for Specifications 3.4.13, RCS [Reactor 

Coolant System] Operational LEAKAGE, 

Specification 5.5.9, Steam Generator 

(SG) Tube Surveillance Program, and 

Specification 5.6.7, Steam Generator 

Tube Surveillance Reports, and add a 

new Specification 3.4.16 entitled Steam 

Generator (SG) Tube Integrity. Also, as 

a result of the licensee replacing the SGs 

with SGs having a new Alloy 690 

thermally treated tubing design, the 

Technical Specifications and Bases 

would be revised to reflect this 

replacement. 

64988 Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 

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 requires a Steam 

Generator Program that includes performance 

criteria that will provide reasonable 

assurance that the steam generator (SG) 

tubing will retain integrity over the full range 

of design basis operating conditions 

(including startup, power operation, hot 

standby, cooldown, anticipated transients 

and postulated accidents). The SG 

performance criteria are based on tube 

structural integrity, accident induced 

leakage, and operational LEAKAGE. These 

criteria assure that the probability of an 

accident will not be increased. 

The primary to secondary accident 

induced leakage rate for any design basis 

accidents, other than an SG tube rupture, 

shall not exceed the leakage rate assumed in 

the accident analysis in terms of total leakage 

rate for all SGs and leakage rate for an 

individual SG. [The primary to secondary 

accident induced leakage rate is relatively 

inconsequential for the SG tube rupture 

analysis.] The operational LEAKAGE 

performance criterion meets current NRC 

regulations and NEI [Nuclear Energy 

Institute] 97–06 criteria for reactor coolant 

system (RCS) operational primary to 

secondary LEAKAGE through any one SG of 

150 gallons per day. These criteria assure that 

accident doses will stay within regulatory 

and licensing basis limits. 

Therefore, the proposed change does not 

affect the probability or consequences of any 

ANO–1 [Arkansas Nuclear One, Unit 1] 

analyzed accidents. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed performance based 

requirements are an improvement over the 

requirements imposed by the current 

technical specifications. Implementation of 

the proposed Steam Generator Program will 

not introduce any adverse changes to the 

plant design basis or postulated accidents 

resulting from potential tube degradation. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary or secondary coolant chemistry 

controls. The proposed change enhances SG 

inspection requirements. 

Therefore, the proposed change does not 

create the possibility of a new or different 

type of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

Steam generator tube integrity is a function 

of the design, environment, and the physical 

condition of the tube. The proposed change 

does not affect tube design or operating 

environment. The proposed change is 

expected to result in an improvement in the 

tube integrity by implementing the Steam 

Generator Program to manage SG tube 

inspection, assessment, repair, and plugging. 

The requirements established by the Steam 

Generator Program are consistent with those 

in the applicable design codes and standards 

and are an improvement over the 

requirements in the current technical 

specifications. 

Therefore, the margin of safety is not 

changed by the proposed change to the 

ANO–1 TSs.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Nicholas S. 

Reynolds, Esquire, Winston and Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Michael K. Webb, 

Acting. 

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request:

September 30, 2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification (TS) 4.2.1, Fuel 

Assemblies, to permit the use of M5 

advanced alloy for fuel rod cladding and 

fuel assembly structural components. 

Also, the proposed amendment would 

modify TS 2.1.1.2, Reactor Core Safety 

Limits, to allow the use of the high 

thermal power (BHTP) correlation for 

departure from nucleate boiling (DNB) 

calculations of reload cores containing 

the Mark-B-HTP fuel design. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The NRC approved topical reports BAW–

10227P–A, Evaluation of Advanced Cladding 

and Structural Material (M5) in PWR 

[Pressurized Water Reactor] Reactor Fuel, 

and BAW–10179P–A, Safety Criteria and 

Methodology for Acceptable Cycle Reload 

Analyses, provide the licensing basis for the 

Framatome ANP (FRA–ANP) advanced 

cladding and structural material, designated 

M5. The M5 material was shown in these 

documents to have equivalent or superior 

properties to the currently used Zircaloy-4 

material. The cladding itself is not an 

accident initiator and does not affect accident 

probability. The M5 cladding has been 

shown to meet all 10 CFR 50.46 design 

criteria and, therefore, will not increase the 

consequences of an accident. 

The proposed safety limit value ensures 

that fuel integrity will be maintained during 

normal operations and anticipated 

operational occurrences (AOOs), and that the 

design requirements will continue to be met. 

The core operating limits will be developed 

in accordance with the new methodology. 

The proposed safety limit value does not 

affect the performance of any equipment 

used to mitigate the consequences of an 

analyzed accident. There is no impact on the 

source term or pathways assumed in 

accidents previously evaluated. No analysis 

assumptions are violated and there are no 

adverse effects on the factors that contribute 

to offsite or onsite dose as the result of an 

accident. 

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. 

Use of M5 clad fuel will not result in 

changes in the operation or configuration of 

the facility. Topical report BAW–10227P–A 

demonstrated that the material properties of 

the M5 alloy are similar or better than those 

of Zircaloy-4. Therefore, M5 fuel rod 

cladding and fuel assembly structural 

components will perform similarly to those 

fabricated from Zircaloy-4, thus precluding 

the possibility of the fuel becoming an 

accident initiator and causing a new or 

different type of accident. 

In addition, there will be no change in the 

level of controls or methodology used for 

processing radioactive effluents or handling 

solid radioactive waste. Since the material 

properties of M5 alloy are similar or better 

than those of Zircaloy-4, there will be no 

significant changes in the types of any 

effluents that may be released off-site. There 

will not be a significant increase in 

occupational or public radiation exposure.

The proposed safety limit value does not 

change the methods governing normal plant 

operation, nor are the methods utilized to 

respond to plant transients altered. The 

BHTP correlation is not an accident / event 

initiator. No new initiating events or 

transients result from the use of the BHTP 

correlation or the related safety limit 

changes. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change will not involve a 

significant reduction in the margin of safety 

because it has been demonstrated that the 

material properties of the M5 alloy are not 

Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 64989

significantly different from those of Zircaloy4. M5 alloy is expected to perform similarly 

or better than Zircaloy-4 for all normal 

operating and accident scenarios, including 

both loss of coolant accident (LOCA) and 

non-LOCA scenarios. For LOCA scenarios, 

where the slight difference in M5 material 

properties relative to Zircaloy-4 could have 

some impact on the overall accident scenario, 

plant-specific LOCA analyses will be 

performed prior to the use of fuel assemblies 

with fuel rods or fuel assembly components 

containing M5. These LOCA analyses, 

required by the ANO–1 [Arkansas Nuclear 

One, Unit 1] TSs, will demonstrate that all 

applicable margins of safety will be 

maintained by the use of M5 alloy. 

The proposed safety limit value has been 

established in accordance with the 

methodology for the BHTP correlation, to 

ensure that the applicable margin of safety is 

maintained (i.e., there is at least 95% 

probability at a 95% confidence level that the 

hot fuel rod in the core does not experience 

DNB). The other reactor core safety limits 

will continue to be met by analyzing the 

reload for the mixed core using NRC 

approved methods, and incorporation of 

resultant operating limits into the Core 

Operating Limits Report (COLR). 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Nicholas S. 

Reynolds, Esquire, Winston and Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Michael K. Webb, 

Acting. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request:

September 1, 2004. 

Description of amendment request:

The proposed amendment would delete 

Technical Specification (TS) 5.6.1, 

‘‘Occupational Radiation Exposure 

Report,’’ and TS 5.6.4, ‘‘Monthly 

Operating Reports.’’

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in license 

amendment applications in the Federal 

Register on June 23, 2004 (69 FR 35067). 

The licensee affirmed the applicability 

of the model NSHC determination in its 

application dated September 1, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated? 

The proposed change eliminates the TS 

reporting requirements to provide a monthly 

operating report of shutdown experience and 

operating statistics if the equivalent data is 

submitted using an industry electronic 

database. It also eliminates the Technical 

Specification reporting requirement for an 

annual occupational radiation exposure 

report, which provides information beyond 

that specified in NRC regulations. The 

proposed change involves no changes to 

plant systems or accident analyses. As such, 

the change is administrative in nature and 

does not affect initiators of analyzed events 

or assumed mitigation of accidents or 

transients. Therefore, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated? 

The proposed change does not involve a 

physical alteration of the plant, add any new 

equipment, or require any existing 

equipment to be operated in a manner 

different from the present design. Therefore, 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety? 

This is an administrative change to 

reporting requirements of plant operating 

information and occupational radiation 

exposure data, and has no effect on plant 

equipment, operating practices or safety 

analyses assumptions. For these reasons, the 

proposed change does not involve a 

significant reduction in the margin of safety.

Based upon the reasoning presented 

above, the requested change does not 

involve significance hazards 

consideration. 

Attorney for licensee: J. M. Fulton, 

Esquire, Assistant General Counsel, 

Pilgrim Nuclear Power Station, 600 

Rocky Hill Road, Plymouth, 

Massachusetts, 02360–5599. 

NRC Section Chief: James W. Clifford. 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota; Docket No. 50–331, Duane 

Arnold Energy Center, Linn County, 

Iowa; Docket No. 50–305, Kewaunee 

Nuclear Power Plant, Kewaunee County, 

Wisconsin; Docket No. 50–255, 

Palisades Plant, Van Buren County, 

Michigan; Docket Nos. 50–266 and 50–

301, Point Beach Nuclear Plant, Units 1 

and 2, Town of Two Creeks, Manitowoc 

County, Wisconsin; Docket Nos. 50–282 

and 50–306, Prairie Island Nuclear 

Generating Plant, Units 1 and 2, 

Goodhue County, Minnesota 

Date of amendment request: October 

5, 2004. 

Description of amendment request:

The proposed amendment would delete 

Technical Specification (TS) 

requirements for the licensee to submit 

annual occupational radiation exposure 

reports and monthly operating reports 

for the above nuclear plants. For the 

Kewaunee and Monticello plants, the 

licensee is also proposing to adopt a 

part of Revision 4 to TSTF–258, 

‘‘Changes to Section 5.0, Administrative 

Controls,’’ regarding reporting 

challenges to, and failures, of certain 

safety/relief valves. 

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in license 

amendment applications in the Federal 

Register on June 23, 2004 (69 FR 35067). 

The licensee affirmed the applicability 

of the model NSHC determination in its 

application dated October 5, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated? 

The proposed change eliminates the TS 

reporting requirements to provide a monthly 

operating report of shutdown experience and 

operating statistics if the equivalent data is 

submitted using an industry electronic 

database. It also eliminates the Technical 

Specification reporting requirement for an 

annual occupational radiation exposure 

report, which provides information beyond 

that specified in NRC regulations. The 

proposed change involves no changes to 

plant systems or accident analyses. As such, 

the change is administrative in nature and 

does not affect initiators of analyzed events 

or assumed mitigation of accidents or 

transients. Therefore, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

64990 Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated? 

The proposed change does not involve a 

physical alteration of the plant, add any new 

equipment, or require any existing 

equipment to be operated in a manner 

different from the present design. Therefore, 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety? 

This is an administrative change to 

reporting requirements of plant operating 

information and occupational radiation 

exposure data, and has no effect on plant 

equipment, operating practices or safety 

analyses assumptions. For these reasons, the 

proposed change does not involve a 

significant reduction in the margin of safety.

Based upon the reasoning presented 

above, the requested change does not 

involve significant hazards 

consideration. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & 

Secretary, Nuclear Management 

Company, LLC, 700 First Street, 

Hudson, WI 54016. 

NRC Section Chief: L. Raghavan. 

Southern Nuclear Operating Company 

(SNC), Inc., et al., Docket Nos. 50–424 

and 50–425, Vogtle Electric Generating 

Plant (VEGP), Units 1 and 2, Burke 

County, Georgia 

Date of amendment request: August 

13, 2004. 

Description of amendment request:

The proposed amendments would 

revise Technical Specification (TS) 

3.7.18, ‘‘Fuel Assembly Storage in the 

Fuel Storage Pool;’’ TS 4.3.1.1, the 

criticality design features for fuel 

storage for VEGP Unit 1; and TS 4.3.1.2, 

the criticality design features for fuel 

storage for VEGP Unit 2. The proposed 

amendment would supplant the 

previous spent fuel rack criticality 

analysis with updated criticality 

calculations. Editorial revisions to TS 

Bases B 3.7.17, ‘‘Fuel Storage Pool 

Boron Concentration,’’ and B 3.7.18, 

‘‘Fuel Assembly Storage in the Fuel 

Storage Pool,’’ are included. In addition, 

Page vi of the Table of Contents will be 

updated to reflect the correct page 

number for Figure 5.5.6–1.

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequence of an accident previously 

evaluated? 

SNC has chosen to reanalyze the criticality 

analyses for the VEGP Unit 1 and Unit 2 

spent fuel racks. Westinghouse performed the 

revised analyses using methods that address 

the non-conservatisms previously identified 

in the current analyses. The methodologies 

used for the revised analysis have been 

previously approved for use by the NRC. 

The analyses revised the enrichment, 

burnup, and Integral Fuel Burnable Absorber 

(IFBA) limits required to comply with the 

allowed storage configurations. The storage 

configurations and interface requirements in 

the current Technical Specifications were 

retained in the revised analyses. The boron 

dilution evaluation that supported the initial 

amendments to permit credit for the soluble 

boron at VEGP continues to remain valid. 

The analyses demonstrated that Keff remains 

below unity for the various storage 

configurations considered with zero soluble 

boron and that Keff remains less than or 

equal to 0.95 for the entire pool with credit 

for soluble boron under non-accident and 

accident conditions with a 95% probability 

at a 95% confidence level (95/95). 

Core design procedures ensure that new 

fuel can be stored in one or more of the 

allowed storage configurations. 

Administrative controls during fuel 

fabrication ensure that the fuel is fabricated 

accordingly to ensure proper loading of the 

fuel in the fuel assemblies. Administrative 

controls used to load fuel assemblies into the 

spent fuel pool ensure that fuel assemblies 

are stored in compliance with the allowed 

storage configurations. Fuel handling is 

performed under many administrative 

controls and physical limitations. These 

controls provide reasonable assurance that a 

criticality accident, fuel fabrication error, or 

fuel handling accident will not occur. 

The change to the page number of Figure 

5.5.6–1 on Page vi of the Table of Contents 

is administrative in nature. 

Therefore, based on the conclusions of the 

above analysis, the proposed change does not 

involve a significant increase in the 

probability or consequence of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

The types of accidents previously 

evaluated include fuel fabrication errors, 

criticality accidents, and fuel handling 

accidents. The analyses revised the 

enrichment, burnup, and Integral Fuel 

Burnable Absorber (IFBA) limits required to 

comply with the allowed storage 

configurations. No new or other kind of 

accident can be postulated as a result of the 

revised analyses. 

The change to the page number of Figure 

5.5.6–1 on Page vi of the Table of Contents 

is administrative in nature. 

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 decrease in the margin of safety? 

The analyses revised the enrichment, 

burnup, and Integral Fuel Burnable Absorber 

(IFBA) limits required to comply with the 

allowed storage configurations. The boron 

dilution evaluation that supported the initial 

amendments to permit credit for soluble 

boron at VEGP was shown to remain valid. 

The analyses demonstrated that Keff remains 

below unity for the various storage 

configurations considered with zero soluble 

boron and that Keff remains less than or 

equal to 0.95 for the entire pool with credit 

for soluble boron under non-accident and 

accident conditions with a 95% probability 

at a 95% confidence level (95/95). 

The change to the page number of Figure 

5.5.6–1 on Page vi of the Table of Contents 

is administrative in nature. 

Therefore, the proposed change does not 

involve a significant decrease 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: Mary Jane RossLee, Acting. 

Tennessee Valley Authority, Docket No. 

50–259, Browns Ferry Nuclear Plant, 

Unit 1, Limestone County, Alabama 

Date of amendment request: July 8, 

2004 (TS–427). 

Description of amendment request:

The proposed amendment removes the 

requirement to maintain an automatic 

transfer capability for the power supply 

to the Low Pressure Coolant Injection 

(LPCI) inboard injection and 

recirculation pump discharge valves. In 

addition, the licensee has requested to 

delete the references to Reactor Motor 

Operator Valve Boards D and E from 

Limiting Condition for Operation 3.8.7, 

and the Actions in 3.8.7 have been 

requested to be revised and/or 

renumbered, as appropriate.

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 Technical 

Specification change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

Neither Reactor Motor Operated Valve 

(RMOV) Boards D and E, the equipment they 

power, nor the automatic power transfer 

Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 64991

feature provided for these boards are 

precursors to any accident previous [sic] 

evaluated in the Updated Final Safety 

Analysis Report (UFSAR). Therefore, the 

probability of an evaluated accident is not 

increased by modifying this equipment. 

The proposed deletion of the requirement 

to maintain an automatic transfer capability 

for the power supply to the LPCI inboard 

injection and recirculation pump discharge 

valves does not change the number of 

Emergency Core Cooling System (ECCS) 

subsystems credited in the BFN licensing 

basis. Therefore, the proposed TS changes 

will not significantly increase the 

consequences of an accident previously 

evaluated. 

2. Does the proposed Technical 

Specification change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed deletion of the requirement 

to maintain an automatic transfer capability 

for the power supply to the LPCI inboard 

injection and recirculation pump discharge 

valves does not introduce new equipment, 

which could create a new or different kind 

of accident. No new external threats, release 

pathways, or equipment failure modes are 

created. Therefore, the proposed deletion of 

the requirement to maintain an automatic 

transfer capability for the power supply to 

the LPCI inboard injection and recirculation 

pump discharge valves will not create a 

possibility for an accident of a new or 

different type than those previously 

evaluated. 

3. Does the proposed Technical 

Specification change involve a significant 

reduction in a margin of safety? 

Response: No. 

The proposed deletion of the requirement 

to maintain an automatic transfer capability 

for the power supply to the LPCI inboard 

injection and recirculation pump discharge 

valves does not change the number of ECCS 

subsystems credited in the BFN licensing 

basis. The requirements of 10 CFR 50.46 and 

Appendix K continue to be met. Therefore, 

the proposed change does not involve a 

significant reduction in the margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Tennessee Valley Authority, Docket No. 

50–259, Browns Ferry Nuclear Plant 

(BFN), Unit 1, Limestone County, 

Alabama 

Date of amendment request: August 2, 

2004 (TS–435). 

Description of amendment request:

Modify the COMPLETION TIME for 

Technical Specification Limiting 

Condition for Operation (LCO) 3.6.3.1, 

Containment Atmosphere Dilution 

(CAD) System. The proposed change 

would extend the current completion 

time of 7 days with two CAD 

subsystems inoperable from existing 

requirement to shut down the reactor 

within 13 hours in accordance with 

LCO 3.0.3, when both CAD subsystems 

are inoperable. 

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 safety-related function of the CAD 

system is to mitigate the effects of a loss-ofcoolant-accident (LOCA) by limiting the 

volumetric concentration of oxygen in the 

primary containment atmosphere. The CAD 

System is not an event initiator, therefore, the 

probability of the occurrence of an accident 

is not affected by this proposed Technical 

Specification change. Emergency procedures 

preferentially use the normal containment 

inerting system to provide post accident vent 

and purge capability, with the CAD system 

only serving in a backup role to this system. 

Hence, in the event of the inoperability of 

both CAD subsystems, the proposed TS 

require the normal containment inerting 

system to be verified available as an alternate 

oxygen control means. Therefore, the 

proposed TS 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 introduce 

new equipment, which could create a new or 

different kind of accident. This proposed 

change does not result in any changes to the 

CAD equipment design or capabilities or to 

the operation of the plant. No new external 

threats, release pathways, or equipment 

failure modes are created. Therefore, the 

implementation of the proposed change will 

not create a possibility for an accident of a 

new or different type than those previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

As stated in GL [Generic Letter] 84–09, a 

Mark I type boiling water reactor (BWR) plant 

does not rely upon purge/repressurization 

systems such as CAD as its primary means 

of hydrogen control when the unit is 

operated in accordance with certain technical 

criteria. The BFN units are operated in 

accordance with these criteria. The BFN Unit 

1 containment is inerted with nitrogen 

during normal operation, nitrogen from the 

containment inerting system with a backup 

from the CAD system is used for 

pneumatically operated components inside 

containment, and there are no potential 

sources of oxygen generation inside 

containment other than the radiolytic 

decomposition of water. The system 

preferred by the Emergency Operating 

Instructions (EOIs) for oxygen control postaccident is the normal primary containment 

inerting system. Because the probability of an 

accident involving hydrogen and oxygen 

production is small, CAD is not the primary 

system used to mitigate the creation of 

combustible containment atmosphere 

mixtures, and because the requested LCO 

where both CAD subsystems is inoperable is 

not long, no significant reduction in the 

margin of safety is associated with this 

proposed amendment.

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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Tennessee Valley Authority (TVA), 

Docket Nos. 50–327 and 50–328, 

Sequoyah Nuclear Plant, Units 1 and 2, 

Hamilton County, Tennessee 

Date of amendment request: August 

18, 2004. 

Description of amendment request:

The proposed amendment would 

update the reactor coolant system (RCS) 

and emergency core cooling system 

(ECCS) technical specifications (TSs). 

These changes include deleting TS 3/

4.4.2, ‘‘Safety Valves—Shutdown’’ in its 

entirety, revising the action 

requirements for TS 3/4.4.3, ‘‘Safety and 

Relief Valves—Operating,’’ and deleting 

surveillance requirement 4.4.3.2.1.a for 

TS 3.4.3.2, ‘‘Relief Valves—Operating.’’

The proposed changes are consistent 

with the Sequoyah (SQN) safety 

analyses provided in the SQN Updated 

Final Safety Analyses Report and the 

improved standard technical 

specifications (NUREG–1431, Revision 

3). 

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:

64992 Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. TVA’s proposed TS revisions do not 

involve a significant increase in the 

probability of any accidents previously 

evaluated. TVA’s proposed TS revisions 

provide improvements to the RCS and ECCS 

requirements to include appropriate 

reference to SQN’s PTLR 

[PressureTemperature Limits Report] 

requirements. The proposed revision is a TS 

improvement that remains consistent with 

the improved standard TS requirements for 

Pressurized Water Reactors (PWRs) (NUREG–

1431, Revision 3). TVA’s proposed revision 

to delete SQN TS 3/4.4.2.1, ‘‘Reactor Coolant 

Safety Valves—Shutdown,’’ does not involve 

a significant increase in the probability of 

any accident previously evaluated. 

Pressurizer code safety valve requirements 

are not applicable for plant shutdown 

conditions (i.e., modes 4 and 5) because the 

valves do not perform a safety function in 

these modes. The pressurizer code safety 

valves are not used as inputs to initiating 

events or accidents previously evaluated. 

Protection of the RCS against an overpressure 

condition in modes 4 and 5 is provided by 

the LTOP [low temperature overpressure 

protection] system which is governed by 

SQN TS 3.4.12. The setpoint for the 

pressurizer code safety valves is sufficiently 

high such that the safety valves do not afford 

protection to the RCS during low temperature 

operation. Accordingly, there is no impact on 

the consequences previously evaluated for 

the proposed change. 

The proposed revisions are not the result 

of changes to plant equipment, test methods 

or operating practices. The proposed changes 

do not contribute to the generation or 

assumptions for postulated accidents. The 

proposed changes do not affect the design 

basis accidents or their assumptions. The 

revisions to SQN TSs continue to support 

SQN’s required safety functions. 

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? 

No. The proposed revisions are not the 

result of changes to plant equipment or plant 

design. The proposed revisions adopt 

standard TS requirements that are consistent 

with SQN’s safety analysis and design and 

provide improvements over the existing 

requirements. The safety functions of the 

RCS and ECCS remain unchanged and do not 

affect any assumptions in SQN’s accident 

analyses. 

TVA’s proposed change to delete the mode 

4 and mode 5 TS requirements for 

pressurizer safety valves is consistent with 

the Policy Criterion of 10 CFR 50.36. The 

pressurizer code safety valves are not 

assumed to function for any safety analysis 

in modes 4 and 5 and consequently, the 

proposed changes do not create the 

possibility of a new or different kind of 

accident. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The proposed TS change does not 

involve a significant reduction in a margin of 

safety. TVA’s proposed revisions will not 

result in changes to system design features or 

plant features that could be precursors to 

accidents or potential degradation of accident 

mitigation systems. The proposed changes to 

the RCS and ECCS requirements remain 

consistent with the current TS requirements 

for equipment operability. Therefore, the 

proposed changes do not involve a 

significant reduction in the margin of safety.

TVA’s proposed change that removes the 

requirement for a pressurizer safety valve in 

modes 4 and 5 does not affect any margin of 

safety because the lift setting of the 

pressurizer code safety valves (2485 pounds 

per square inch gauge [psig] ±3 percent) is 

well above the limit needed to protect the 

RCS during low temperature operation and 

would not provide any safety function for 

overpressure protection in the lower modes. 

The TS requirements associated with low 

temperature operation are governed by SQN 

TS 3/4.4.12, LTOP system. The LTOP system 

provides the necessary overpressure 

protection for SQN’s RCS in modes 4 and 5. 

Accordingly, TVA’s proposed deletion of 

operability requirements for SQN’s 

pressurizer code safety valves for modes 4 

and 5 will not affect the margin of safety.

The United States Nuclear Regulatory 

Commission (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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request: October 

12, 2004. 

Brief description of amendment 

request: The proposed amendment 

would approve an engineering 

evaluation performed in accordance 

with Pilgrim Nuclear Power Station 

Technical Specification (TS) 3.6.D.3 to 

justify continued power operation with 

a safety relief valve discharge pipe 

temperature exceeding 212 degrees 

Fahrenheit for greater than 24 hours as 

required by TS 3.6.D.4. 

Date of publication of individual 

notice in Federal Register: October 20, 

2004 (69 FR 61695). 

Expiration date of individual notice:

December 19, 2004. 

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 

Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 64993

items are available for public inspection 

at the Commission’s Public Document 

Room, 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 NRC 

Public Document Room (PDR) Reference 

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

or by e-mail to pdr@nrc.gov. (Note: 

Public access to ADAMS has been 

temporarily suspended so that security 

reviews of publicly available documents 

may be performed and potentially 

sensitive information removed. Please 

check the NRC Web site for updates on 

the resumption of ADAMS access.) 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of application for amendment:

July 30, 2004. 

Brief description of amendment: The 

proposed amendment would (1) add 

License Condition 2.C.(22) requiring an 

integrated tracer gas test of the control 

room envelope using methods described 

in American Society for Testing and 

Materials E741–00, ‘‘Standard Test 

Method for Determining Air Change in 

a Single Zone by Means of a Tracer Gas 

Dilution,’’ and (2) delete Surveillance 

Requirement 3.7.3.6, which requires 

verification that unfiltered inleakage 

from control room emergency filtration 

system duct work outside the control 

room envelope is within limits. 

Date of issuance: October 25, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 162. 

Facility Operating License No. NPF–

43: Amendment adds a license 

condition and revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 13, 2004 (69 FR 50217) 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 25, 

2004. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States, Inc., and Entergy 

Operations, Inc., Docket No. 50–458, 

River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: October 

21, 2003, as supplemented by letters 

dated February 10, 2004, and August 24, 

2004.

Brief description of amendment:

Modifies the Technical Specifications 

(TSs) to delete TS 3.6.4.4, ‘‘Shield 

Building Annulus Mixing System’’ and 

a reference to TS 3.6.4.4 within TS 

3.10.1, ‘‘Inservice Leak and Hydrostatic 

Testing Operation,’’ and revise TS 

Surveillance Requirement 3.6.1.3.10, 

main steam isolation valve leakage 

limits. 

Date of issuance: October 15, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 143. 

Facility Operating License No. NPF–

47: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 25, 2004 (69 FR 29764). The 

supplement dated August 24, 2004, 

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 15, 

2004. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States, Inc., and Entergy 

Operations, Inc., Docket No. 50–458, 

River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: February 

16, 2004, as supplemented by letters 

dated June 8 and August 26, 2004. 

Brief description of amendment:

Modifies the Technical Specifications 

(TSs) to change Surveillance 

Requirement 3.6.5.1.3 of TS 3.6.5.1, 

‘‘Drywell,’’ to allow a one-time 

extension of the test interval for the next 

drywell bypass leakage rate test from 10 

years to 15 years. 

Date of issuance: October 15, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 144. 

Facility Operating License No. NPF–

47: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 25, 2004 (69 FR 29765). The 

supplements dated June 8 and August 

26, 2004, 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 15, 

2004. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Docket No. 

50–247, Indian Point Nuclear 

Generating Unit No. 2, Westchester 

County, New York 

Date of application for amendment:

January 29, 2004, as supplemented on 

April 12, June 16, June 30, July 16, 

August 3, August 12, and September 24, 

2004. 

Brief description of amendment: The 

amendment revises the operating 

license and Technical Specifications to 

authorize an increase in the maximum 

steady-state reactor core power level 

from 3114.4 megawatt thermal (MWt) to 

3216 MWt. This represents a nominal 

increase of 3.26% rated thermal power. 

Date of issuance: October 27, 2004. 

Effective date: As of the date of 

issuance to be implemented within 30 

days. 

Amendment No.: 241. 

Facility Operating License No. DPR–

26: Amendment revised the Facility 

Operating License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 2, 2004 (69 FR 9859). The 

April 12, June 16, July 16, August 3, 

August 12, and September 24, 2004, 

supplements provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 27, 

2004. 

No significant hazards consideration 

comments received: No. 

64994 Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of application for amendments:

January 15, 2004, and supplemented on 

July 19, 2004. 

Brief description of amendments: The 

amendments provide for an alternative 

means of testing the main steam 

Electromatic relief valves and the dual 

function Target Rock safety/relief 

valves. 

Date of issuance: October 19, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 211/203, 222/217. 

Facility Operating License Nos. DPR–

19, DPR–25, DPR–29 and DPR–30: The 

amendments revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 16, 2004. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 19, 

2004. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of application for amendments:

August 19, 2003. 

Brief description of amendments: The 

amendments modify Technical 

Specification (TS) 5.5.13, ‘‘Primary 

Containment Leakage Rate Testing 

Program,’’ to allow an exception to the 

testing guidance contained in 

Regulatory Guide 1.163, ‘‘PerformanceBased Containment Leak-Test Program.’’

Specifically, the TS change will allow 

potential valve atmospheric leakage 

paths (e.g., valve stem packing) that are 

not exposed to test pressure during 

reverse-direction Type B or C tests (local 

leakage rate tests) to instead be tested 

during regularly scheduled Type A tests 

(integrated leakage rate tests). 

Date of issuance: October 14, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 168/154. 

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: December 23, 2003 (68 FR 

74266). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 14, 

2004.

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–334, 

Beaver Valley Power Station, Unit No. 1 

(BVPS–1), Beaver County, Pennsylvania 

Date of application for amendment:

June 28, 2004, as supplemented 

September 3, 2004. 

Brief description of amendment: The 

amendment revised the BVPS–1 

Technical Specifications (TSs) 

surveillance requirements (SRs) 

4.4.5.4.a.6, 4.4.5.4.a.8, and 4.4.5.5.d.1 

and added SRs 4.4.5.4.a.11 and 4.4.5.5.e 

for Cycle 17 operation only. The change 

revised the definition of steam generator 

tube inspection scope in SR 4.4.5.4.a.8 

to exclude the portion of the tube within 

the tubesheet below the W* distance, 

tube to tubesheet weld and tube-end 

extension by crediting the Westinghouse 

W* methodology as described in 

Topical Report WCAP–14797, Revision 

2. 

Date of issuance: October 15, 2004. 

Effective date: This license 

amendment is effective as of its date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 262. 

Facility Operating License No. DPR–

66: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 3, 2004 (69 FR 46584). 

The supplement dated September 3, 

2004, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the Nuclear Regulatory Commission 

staff’s original proposed no significant 

hazards consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 15, 

2004. 

No significant hazards consideration 

comments received: No. 

Florida Power Corporation, et al., 

Docket No. 50–302, Crystal River Unit 

No. 3 Nuclear Generating Plant, Citrus 

County, Florida 

Date of application for amendment:

December 9, 2003, as supplemented 

September 16, 2004. 

Brief description of amendment: The 

amendment allows a one-time increase 

in the completion time for restoring an 

inoperable emergency feedwater (EFW) 

system train to operable status to allow 

the realignment of the diesel-driven 

EFW pump during power operations. 

Date of issuance: October 21, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 214. 

Facility Operating License No. DPR–

72: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 30, 2004 (69 FR 

16620). The September 16, 2004, 

supplemental letter provided additional 

information that clarified the 

application, but did not expand the 

scope of the application as originally 

noticed and did not change the U.S. 

Nuclear Regulatory Commission staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 21, 

2004. 

No significant hazards consideration 

comments received: No. 

FPL Energy Seabrook, LLC, Docket No. 

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

Rockingham County, New Hampshire 

Date of amendment request: February 

3, 2004. 

Description of amendment request:

This amendment revised a footnote to 

clarify a surveillance requirement and 

associated bases for emergency diesel 

generator testing. 

Date of issuance: October 25, 2004. 

Effective date: As of its date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 98. 

Facility Operating License No. NPF–

86: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 16, 2004 (69 FR 

12371). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 25, 

2004. 

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: July 15, 

2004, as supplement by letters dated 

September 28 and October 14, 2004. 

Brief description of amendment: The 

amendment revises the Technical 

Specification (TS) Section 3.8.1, AC 

Sources—Operating, Condition B, to 

provide a one-time extension of the 

allowed outage time for one Diesel 

Generator (DG) inoperable from 7 days 

Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 64995

to 14 days and TS Section 3.8.3, Diesel 

Fuel Oil, Lube Oil, and Starting Air, 

Limiting Condition for Operation, to 

allow the use of temporary fuel oil 

storage tanks to supply the required fuel 

oil storage inventory. 

Date of issuance: October 15, 2004. 

Effective date: As of the date of 

issuance and shall be implemented on 

or before October 22, 2004. 

Amendment No.: 207.

Facility Operating License No. DPR–

46: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 3, 2004 (69 FR 46586). 

The supplements dated September 28 

and October 14, 2004, 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 15, 

2004. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket Nos. 50–282 and 50–306, Prairie 

Island Nuclear Generating Plant, Units 

1 and 2, Goodhue County, Minnesota 

Date of application for amendments:

December 23, 2003, as supplemented by 

letter dated August 16, 2004. 

Brief description of amendments: The 

amendments modify 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, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days. 

Amendment Nos.: 167, 157. 

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: September 16, 2004 (69 FR 

55844) The supplement dated August 

16, 2004, 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 20, 

2004. 

No significant hazards consideration 

comments received: No. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request:

December 1, 2003, as supplemented by 

letter dated July 2, 2004. 

Brief description of amendment: The 

amendment changes the Fort Calhoun 

Station, Unit No. 1 Technical 

Specifications (TS) 2.7, ‘‘Electrical 

Systems, TS Table 3–5, ‘‘Minimum 

Frequencies for Equipment Tests,’’ and 

TS 5.0, ‘‘Administrative Controls,’’ to 

modify the requirements for the diesel 

generator (DG) fuel oil for consistency 

with the Improved Standard Technical 

Specifications. The amendment also 

adds requirements for the DG 

lubricating oil and DG starting air. 

Date of issuance: October 21, 2004. 

Effective date: October 21, 2004, and 

shall be implemented within 120 days 

from the date of its issuance. 

Amendment No.: 229. 

Renewed Facility Operating License 

No. DPR–40: The amendment revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: February 17, 2004 (69 FR 7526). 

The additional information provided in 

the supplemental letter dated July 2, 

2004, did not expand the scope of the 

application as noticed and did not 

change the NRC staff’s original proposed 

no significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

safety evaluation dated October 21, 

2004. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment:

October 23, 2003, as supplemented by 

letters dated June 24, 2004 and August 

26, 2004. 

Brief description of amendment: The 

amendment revised Technical 

Specifications to delete the Surveillance 

Requirement associated with the 

emergency diesel generator lockout 

features. 

Date of issuance: October 22, 2004. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 155. 

Facility Operating License No. NPF–

57: This amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 9, 2003

(68 FR 68671). The June 24, 2004, and 

August 26, 2004, letters provided 

clarifying information that did not 

change the initial proposed no 

significant hazards consideration 

determination or expand the application 

beyond the scope of the original 

Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 22, 

2004. 

No significant hazards consideration 

comments received: No.

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment:

December 12, 2003. 

Brief description of amendment: The 

amendment revised the operating 

conditions for which Technical 

Specification (TS) 3/4.3.7.1, ‘‘Radiation 

Monitoring Instrumentation,’’ requires 

the control room ventilation radiation 

monitor to be operable. Additionally, 

the amendment revised the operating 

conditions for which TS 3/4.7.2, 

‘‘Control Room Emergency Filtration 

System,’’ is applicable. 

Date of issuance: October 28, 2004. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 156. 

Facility Operating License No. NPF–

57: This amendment revised the TSs. 

Date of initial notice in Federal 

Register: February 17, 2004 (69 FR 

7527). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 28, 

2004. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment:

December 24, 2003, as supplemented by 

letter dated June 8, 2004. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications (TSs) to allow the use of 

GE14 fuel in reload cycle 13. 

Specifically, the change modified the 

TSs to reflect the use of General Electric 

(GE) core reload analysis methodology. 

The change revised the limiting 

conditions for operation for the 

recirculation loops to modify and add 

action statements to provide further 

thermal limit control during single-loop 

operation to be consistent with the GE 

methodology specified in the core 

operating limits report. The change also 

64996 Federal Register / Vol. 69, No. 216 / Tuesday, November 9, 2004 / Notices 

modified the TS definitions and TS 

requirements for average planar linear 

heat generation rate. Additionally, TS 

Section 6.9.1.9 is revised to correct an 

error from a previous amendment that 

inadvertently removed a reference. 

Date of issuance: October 20, 2004. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 154. 

Facility Operating License No. NPF–

57: This amendment revised the TSs. 

Date of initial notice in Federal 

Register: February 17, 2004 (69 FR 

7528). The June 8, 2004 letter provided 

clarifying information that did not 

change the initial proposed no 

significant hazards consideration 

determination or expand the application 

beyond the scope of the original Federal 

Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 20, 

2004. 

No significant hazards consideration 

comments received: No.

Dated in Rockville, Maryland, this 1st day 

of November 2004.

For the Nuclear Regulatory Commission. 

Ledyard B. Marsh, 

Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 04–24804 Filed 11–8–04; 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:

Wilkins 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) 415–5788; fax 

number: (301) 415–5370; e-mail: 

wrs@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The Nuclear Regulatory Commission 

(NRC) plans to issue Interim Staff 

Guidance (ISG) documents for fuel cycle 

facilities. These ISG documents provide 

clarifying guidance to the NRC staff 

when reviewing either a license 

application or a license amendment 

request for a fuel cycle facility under 10 

CFR part 70. The NRC is soliciting 

public comments on the ISG documents 

which will be considered in the final 

versions or subsequent revisions. 

II. Summary 

The purpose of this notice is to 

provide the public an opportunity to 

review and comment on a draft Interim 

Staff Guidance document for fuel cycle 

facilities. Interim Staff Guidance–09 

provides guidance to NRC staff relative 

the requirements associated with the 

use of Initiating Event Frequencies 

(IEFs) for demonstrating compliance 

with the performance requirements of 

10 CFR 70.61. 

III. Interim Staff Guidance–09, 

Initiating Event Frequency, Draft 

October 20, 2004 Issue 

This guidance addresses the measures 

needed to assure the validity and 

maintenance of initiating event 

frequencies (IEFs) used to demonstrate 

compliance with the performance 

requirements for 10 CFR 70.61. 

Introduction 

The purpose of this Interim Staff 

Guidance (ISG) is to clarify the use of 

IEFs for demonstrating compliance with 

the performance requirements of 10 CFR 

70.61. NUREG–1718, ‘‘Standard Review 

Plan for the Review of an Application 

for a Mixed Oxide (MOX) Fuel 

Fabrication Facility,’’ and NUREG–

1520, ‘‘Standard Review Plan for the 

Review of a License Application for a 

Fuel Cycle Facility,’’ provide methods 

for reviewing integrated safety analyses 

(ISAs), employing a semi-quantitative 

risk index method. While one of these 

methods is used below to illustrate the 

use of IEFs, applicants and licensees 

may use other methods which would 

produce similar results. There is no 

particular method explicitly mandated, 

and sequences that are risk significant 

or marginally acceptable are candidates 

for more detailed evaluation by the 

applicant or licensee and reviewer. 

Discussion 

Each licensee or applicant is required 

to perform an ISA to identify all 

credible high-consequence and 

intermediate-consequence events. The 

risk of each such credible event is to be 

limited through the use of appropriate 

engineered and/or administrative 

controls to meet the performance 

requirements of 10 CFR 70.61. Such a 

control is referred to as an item relied 

on for safety (IROFS). In turn, a safety 

program must be established and 

maintained to assure that each IROFS is 

available and reliable to perform its 

intended function when needed. The 

safety program may be graded such that 

management measures applied are 

graded commensurate with the 

reduction of risk attributable to that 

item. In addition, a configuration 

management system must be established 

pursuant to § 70.72, to evaluate changes, 

to assure, in part, that the IROFS are not 

removed without at least equivalent 

replacement of the safety function. 

The risk of each credible event is 

determined by cross-referencing the 

severity of the consequence of the 

unmitigated accident sequence with the 

likelihood of occurrence in a risk matrix 

with risk index values. The likelihood 

of occurrence risk index values can be 

determined by considering the criteria 

in NUREG–1520, Tables A–9 through 

A–11. Accident sequences result from 

initiating events which are followed by 

the failure of one or more IROFS. 

Initiating events can be (1) an external 

event such as a hurricane or earthquake, 

(2) a facility event external to the 

process being analyzed (e.g., fires, 

explosions, failures of other equipment, 

flooding from facility water sources), (3) 

deviations from normal operations of 

the process (credible abnormal events), 

or (4) failures of an IROFS in the 

process. Additional guidance regarding 

initiating probabilities from natural 

phenomena hazards are addressed in 

ISG–08, Natural Phenomena Hazards. 

An initiating event does not have to 

be an IROFS failure. An item only 

becomes an IROFS if it is credited in the 

ISA for mitigation or prevention per the 

definition in § 70.4. If an item, whose 

failure initiates an event, has strictly an 

operational function, it does not have to 

be an IROFS. This applies to external 

events and can apply to internal events. 

If the item whose failure initiates an 

event, has solely a safety function that 

is credited in the ISA, then it should be 

an IROFS. If the item has both an 

operational and a safety function, the 

safety function should make it an IROFS 

(for its ISA credited safety features 

only). 

IEFs can play a significant role in 

determining whether the performance 

requirements of § 70.61 are met for a 

particular accident sequence. Whether 

an initiating event is due to an IROFS 

or a non-IROFS failure, licensees should 

take appropriate action to assure that 

any change to the basis for assigning an 

IEF value to that event is evaluated on 

a continuing basis to ensure continued 

compliance with the performance 

requirements. For example, a nonIROFS component may not be subject to 

the same QA program controls and other 

management measures that an IROFS