Federal Register / Vol. 69, No. 194 / Thursday, October 7, 2004 / Notices 60193

Web site to Carol A. Gallagher (301) 

415–5905; email CAG@nrc.gov. 

Hand-deliver comments to: Rules and 

Directives Branch, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, 11555 Rockville Pike, 

Rockville, Maryland 20852, between 

7:30 a.m. and 4:15 p.m. on Federal 

workdays. 

Fax comments to: Rules and 

Directives Branch, Office of 

Administration, U.S. Nuclear Regulatory 

Commission at (301) 415–5144. 

Request for information about draft 

regulatory guide DG–1139 may be 

directed to Paul W. Lain at (301) 415–

2346 or via email to PWL@nrc.gov. 

Comments would be most helpful if 

received by December 15, 2004. 

Comments received after this date will 

be considered if it is practical to do so, 

but the NRC is able to ensure 

consideration only for comments 

received on or before this date. 

Although a time limit is given, 

comments and suggestions in 

connection with items for inclusion in 

guides currently being developed or 

improvements in all published guides 

are encouraged at any time. 

Electronic copies of the draft 

regulatory guide are available through 

the NRC’s public Web site under Draft 

Regulatory Guides in the Regulatory 

Guides document collection of the 

NRC’s Electronic Reading Room at

http://www.nrc.gov/reading-rm/doccollections/. Electronic copies are also 

available in the NRC’s Agencywide 

Documents Access and Management 

System (ADAMS) at http://

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

under Accession No. ML042740308. In 

addition, regulatory guides are available 

for inspection at the NRC’s Public 

Document Room (PDR), which is 

located at 11555 Rockville Pike, 

Rockville, Maryland; the PDR’s mailing 

address is USNRC PDR, Washington, DC 

20555–0001. The PDR can also be 

reached by telephone at (301) 415–4737 

or (800) 397–4205, by fax at (301) 415–

3548; and by email to PDR@nrc.gov. 

Requests for single copies of draft or 

final guides (which may be reproduced) 

or for placement on an automatic 

distribution list for single copies of 

future draft guides in specific divisions 

should be made in writing to the U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attention: 

Reproduction and Distribution Services 

Section; by email to 

DISTRIBUTION@nrc.gov; or by fax to 

(301) 415–2289. Telephone requests 

cannot be accommodated. Regulatory 

guides are not copyrighted, and 

Commission approval is not required to 

reproduce them. (U.S.C. 552(a))

Dated at Rockville, Maryland, this 30th day 

of September , 2004. 

For the U.S. Nuclear Regulatory 

Commission. 

Sharon D. Stewart, 

Acting Director, Program Management, Policy 

Development and Analysis Staff, Office of 

Nuclear Regulatory Research.

[FR Doc. 04–22545 Filed 10–6–04; 8:45 am] 

BILLING CODE 7590–01–P

NUCLEAR REGULATORY 

COMMISSION 

Proposed Generic Communication; 

Steam Generator Tube Integrity and 

Associated Technical Specifications

AGENCY: Nuclear Regulatory 

Commission.

ACTION: Notice of opportunity for public 

comment. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is proposing to issue 

a generic letter (GL) to request that 

addressees submit a description of their 

program for ensuring steam generator 

(SG) tube integrity for the interval 

between inspections and description of 

the methodology used to assess the 

effects of non-pressure-related loads 

such as bending on SG tube integrity. 

Addressees should also provide a safety 

assessment demonstrating that the SG 

tubes will have adequate structural and 

leakage integrity (with appropriate 

regulatory margins) at the time of their 

next SG tube inspection, taking into 

account the effects of non-pressurerelated loads. 

This Federal Register notice is 

available through the NRC’s 

Agencywide Documents Access and 

Management System (ADAMS) under 

accession number ML042710075.

DATES: Comment period expires 

December 6, 2004. Comments submitted 

after this date will be considered if it is 

practical to do so, but assurance of 

consideration cannot be given except for 

comments received on or before this 

date.

ADDRESSEES: Submit written comments 

to the Chief, Rules and Directives 

Branch, Division of Administrative 

Services, Office of Administration, U.S. 

Nuclear Regulatory Commission, Mail 

Stop T6-D59, Washington, DC 20555–

0001, and cite the publication date and 

page number of this Federal Register

notice. Written comments may also be 

delivered to NRC Headquarters, 11545 

Rockville Pike (Room T–6D59), 

Rockville, Maryland, between 7:30 am 

and 4:15 pm on Federal workdays.

FOR FURTHER INFORMATION CONTACT:

Kenneth Karwoski, NRR at 301–415–

2752 or by e-mail at kjk1@nrc.gov or 

Maitri Banerjee, NRR at 301–415–2277 

or by e-mail at mxb@nrc.gov.

SUPPLEMENTARY INFORMATION: 

Draft NRC Generic Letter 2004–XX: 

Steam Generator Tube Integrity and 

Associated Technical Specifications 

Addressees 

All holders of operating licenses for 

pressurized-water reactors (PWRs), 

except those who have permanently 

ceased operations and have certified 

that fuel has been permanently removed 

from the reactor vessel and {the 

following plants that have already 

modified their technical specifications 

to be consistent with those in the 

Attachment}. 

Purpose 

The U.S. Nuclear Regulatory 

Commission (NRC) is issuing this 

generic letter (GL) to: 

(1) Request that addressees submit a 

description of their program for 

ensuring steam generator (SG) tube 

integrity for the interval between 

inspections; and 

(2) Request that addressees submit a 

description of the methodology used to 

assess the effects of non-pressure-related 

loads such as bending on SG tube 

integrity. Addressees should also 

provide a safety assessment 

demonstrating that the SG tubes will 

have adequate structural and leakage 

integrity (with appropriate regulatory 

margins) at the time of their next SG 

tube inspection, taking into account the 

effects of non-pressure-related loads. 

Discussion 

Steam generator tubes function as an 

integral part of the reactor coolant 

pressure boundary (RCPB) and also 

serve to isolate radiological fission 

products in the primary coolant from 

the secondary coolant and the 

environment. For the purposes of this 

generic letter, tube integrity means that 

the tubes are capable of performing 

these functions in accordance with the 

plant design and licensing basis, 

including applicable regulatory 

requirements. 

During operation, licensees are 

required to monitor and maintain the 

condition of the SG tubing with the 

objective of ensuring its continued 

integrity. Specifically, licensees are 

required by 10 CFR 50.55a(b)(2)(iii), 10 

CFR 50.55a(g), or by the plant technical 

specifications to perform periodic 

inservice inspections and to repair (e.g., 

sleeve) or remove from service (by 

installing plugs in the tube ends) all 

tubes found to contain flaws exceeding 

60194 Federal Register / Vol. 69, No. 194 / Thursday, October 7, 2004 / Notices 

1Or, for PWR facilities licensed before the 

issuance of 10 CFR part 50, Appendix A, similar 

requirements in the plant-specific principal design 

criteria.

the plugging limit (i.e., tube repair 

criteria). 

The current technical specification 

requirements for inspection and repair 

of SG tubing were developed in the 

1970s and define a prescriptive 

approach for ensuring tube integrity. 

This prescriptive approach involves 

inspecting the tubing at specified 

intervals, implementing specified tube 

inspection sampling plans, and 

repairing or removing from service by 

plugging all tubes found by inspection 

to contain flaws in excess of specified 

flaw repair criteria. However, as 

evidenced by operating experience, the 

prescriptive approach defined in the 

technical specifications may not be 

sufficient to ensure that tube integrity is 

maintained. For example, in cases of 

low to moderate levels of degradation, 

the technical specifications only require 

that 3-to 21-percent of the tubes be 

inspected, irrespective of whether the 

inspection results indicate that 

additional tubes need to be inspected to 

reasonably ensure that tubes with flaws 

that may exceed the tube repair criteria 

or which may impair tube integrity are 

detected. In addition, the technical 

specifications (and Section XI of the 

American Society of Mechanical 

Engineers (ASME) Boiler and Pressure 

Vessel Code) do not explicitly address 

the inspection methods to be employed 

for different tube degradation 

mechanisms at specific tube locations, 

nor are the specific objectives to be 

fulfilled by the selected methods 

explicitly defined. Also, incremental 

flaw growth between inspections can in 

many instances exceed what is allowed 

for in the specified tube repair criteria. 

In such cases, the specified inspection 

frequencies may not ensure reinspection 

of a tube before its integrity is impaired. 

In short, current technical specification 

surveillance requirements may not 

require licensees to actively manage 

their SG programs so as to provide 

reasonable assurance that tube integrity 

is maintained. As a result of the above, 

licensees have frequently found it 

necessary to implement measures 

beyond the technical specification 

requirements to ensure adequate tube 

integrity. These measures are frequently 

accompanied by interaction with the 

NRC staff in an oversight or review 

capacity to ensure that adequate tube 

integrity is being maintained.

The NRC staff, with external 

stakeholder involvement, embarked on 

efforts to improve the SG tube integrity 

regulatory framework as discussed in 

SECY–03–0800, ‘‘Steam Generator Tube 

Integrity (SGTI)—Plans for Revising the 

Associated Regulatory Framework.’’ As 

a result of these efforts, the NRC and 

industry generically developed 

modified technical specifications for 

addressing steam generator tube 

integrity. These generically developed 

technical specifications were recently 

incorporated into one facility’s technical 

specifications. (Proposals to change the 

plant-specific technical specifications 

are reviewed in accordance with the 

license amendment review process to 

confirm their acceptability). These 

modified technical specifications are 

attached to this generic letter for your 

information. The approach taken in the 

modified technical specifications in the 

Attachment is conceptually similar to 

the approach outlined in the industry 

initiative referred to as NEI 97–06, 

‘‘Steam Generator Program Guidelines.’’

The modified technical specifications in 

the Attachment are performance-based 

in that they are focused on ensuring that 

the tubing satisfies performance criteria 

that are commensurate with tube 

integrity. This approach can be readily 

adapted to new or unexpected 

degradation mechanisms and advances 

in nondestructive examination 

technology. This approach also includes 

programmatic elements to ensure that 

tubes are being adequately monitored 

and maintained relative to the 

performance criteria. 

The requirements pertaining to the 

integrity of the SG tubes are contained 

within Title 10 of the Code of Federal 

Regulations (10 CFR). Specifically, the 

general design criteria (GDC) 1 described 

in Appendix A to 10 CFR part 50 

contain, in part, requirements related to 

the RCPB (e.g., GDC 14, GDC 30, GDC 

32). In addition to the GDC, 10 CFR 

50.55a specifies that components that 

are part of the RCPB must meet the 

requirements for Class 1 components in 

Section III and XI of the American 

Society of Mechanical Engineers 

(ASME) Boiler and Pressure Vessel Code 

unless the plant technical specifications 

for surveillance differ from those 

specified in the ASME Code, in which 

case the technical specifications govern.

The requirements pertaining to the 

content of a plant’s technical 

specifications are given in 10 CFR 50.36, 

‘‘Technical Specifications.’’ All 

currently operating PWR licensees have 

technical specifications governing the 

surveillance of the SG tubes. These 

technical specifications also include 

operational leakage limits so that if 

significant leakage develops, the plant is 

shut down. The plugging limits in the 

technical specifications were developed 

to ensure that degraded tubes: (1) 

Maintain factors of safety against gross 

rupture consistent with the plant design 

basis (i.e., consistent with the stress 

limits of the ASME Code, Section III); 

and (2) maintain leakage integrity 

consistent with the plant licensing basis 

while, at the same time, allowing for 

potential flaw size measurement error 

and flaw growth between inservice 

inspections. 

As part of the plant licensing basis, 

applicants for PWR licenses are also 

required to analyze the consequences of 

postulated design-basis accidents. 

Typical accidents analyzed are the SG 

tube rupture, the locked-rotor, control 

rod ejection, and a main steamline 

break. These analyses consider the 

potential primary-to-secondary leakage 

through the tubes during these events 

and must show that the offsite 

radiological doses do not exceed 10 CFR 

part 100 limits (or some fraction thereof) 

and GDC 19 of Appendix A to 10 CFR 

part 50. 

Irrespective of technical specification 

requirements for SG tube inspection and 

repair, licensees are also required by 10 

CFR part 50, Appendix B Criterion XVI, 

‘‘Corrective Action,’’ to ensure that 

conditions adverse to quality are 

promptly identified and corrected. In 

the case of significant conditions 

adverse to quality, the measures shall 

assure that the cause of the condition is 

determined and corrective action taken 

to preclude repetition. 

The staff is requesting information as 

to: (1) Actions licensees are taking or 

will take to ensure tube integrity is 

being maintained, and (2) contemplated 

changes to the technical specifications 

to reflect these actions. 

As discussed above, the approach in 

the attached technical specifications is 

performance-based. There are three 

performance criteria for the SG tubes: 

(1) A structural integrity performance 

criterion, (2) a primary-to-secondary 

leakage performance criterion for 

normal operation, and (3) a primary-tosecondary leakage performance criterion 

for postulated accident conditions. 

During public interactions with 

stakeholders on the structural integrity 

performance criterion, the staff became 

aware that the effects of various nonpressure-related loads such as bending 

loads may not be fully addressed in 

industry guidance documents for 

assessing the integrity of degraded SG 

tubes. Non-pressure-related loads were 

assessed in the original design of the SG 

tubes so as to ensure that nondegraded 

tubes would have adequate integrity for 

the full range of operating conditions. 

As a result, this generic letter requests 

addressees to discuss how they have 

Federal Register / Vol. 69, No. 194 / Thursday, October 7, 2004 / Notices 60195

assessed the effects of non-pressurerelated loads in their assessments of 

tube integrity and to discuss whether all 

tubes will have adequate structural 

integrity at the time of their next SG 

tube inspection, taking all loading 

conditions on the tube into account.

Requested Information 

Addressees are requested to provide 

the following information to the NRC 

within 60 days of the date of this 

generic letter: 

1. A description of the actions they 

are taking or will take to ensure tube 

integrity is being maintained and 

contemplated changes to the technical 

specifications to reflect these actions. 

2. A description of the methodology 

used to assess the effects of nonpressure-related loads such as bending 

on SG tube integrity. In addition, 

addressees should provide a safety 

assessment demonstrating that the SG 

tubes will have adequate structural and 

leakage integrity at the time of their next 

SG tube inspection, taking into account 

the effects of non-pressure-related loads. 

Required Response 

In accordance with 10 CFR 50.54(f), 

addressees are required to submit 

written responses to this generic letter. 

Two options are available: 

(a) Addressees may choose to submit 

written responses providing the 

information requested above within the 

requested time period. 

(b) Addressees who cannot meet the 

requested completion date or who 

choose an alternate course of action are 

required to so notify the NRC in writing 

as soon as possible but no later than 30 

days from the date of this generic letter. 

The response must address any 

alternative course of action proposed, 

including the basis for the acceptability 

of the proposed alternative course of 

action, and the basis for finding that the 

SGs remain operable. If the information 

requested in the previous section of this 

GL will be subsequently provided, the 

response must set forth the schedule for 

submitting the information. 

The required written response should 

be addressed to the U.S. Nuclear 

Regulatory Commission, ATTN: 

Document Control Desk, 11555 

Rockville Pike, Rockville, Maryland 

20852, under oath or affirmation under 

the provisions of section 182a of the 

Atomic Energy Act of 1954, as amended, 

and 10 CFR 50.54(f). In addition, a copy 

of the response should be sent to the 

appropriate regional administrator. 

Reasons for Requested Information 

This GL requests addressees to submit 

information. The requested information 

will enable the NRC staff to determine 

whether addressees’ SG tube integrity 

programs provide reasonable assurance 

of tube integrity consistent with their 

design and licensing basis and 

applicable regulatory requirements (10 

CFR part 50, Appendix A1; 10 CFR part 

50, Appendix B). In addition, the 

requested information will enable the 

NRC staff to determine whether SG tube 

integrity is being maintained under all 

loading conditions consistent with the 

design and licensing basis and 

applicable regulatory requirements (10 

CFR part 50, Appendix A1). 

The NRC staff will review the 

responses to this GL in order to 

determine whether additional actions 

are necessary. 

Backfit Discussion 

Under the provisions of section 182a 

of the Atomic Energy Act of 1954, as 

amended, and 10 CFR 50.54(f), this GL 

transmits an information request for the 

purpose of verifying compliance with 

applicable existing requirements. 

Specifically, the requested information 

will enable the NRC staff to determine 

whether the applicable requirements 

discussed above are being met. No 

backfit is either intended or approved in 

the context of issuance of this GL. 

Therefore, the staff has not performed a 

backfit analysis. 

Federal Register Notification 

To be done after the public comment 

period. 

Small Business Regulatory Enforcement 

Fairness Act 

The NRC has determined that this 

action is not subject to the Small 

Business Regulatory Enforcement 

Fairness Act of 1996. 

Paperwork Reduction Act Statement 

This GL contains information 

collection requirements that are subject 

to the Paperwork Reduction Act of 1995 

(44 U.S.C. 3501 et seq.). These 

information collections were approved 

by the Office of Management and 

Budget, approval no. 3150–0011, which 

expires on February 28, 2007. 

The burden of these mandatory 

information collections on the public is 

estimated to average 200 hours per 

response, including the time for 

reviewing instructions, searching 

existing data sources, gathering and 

maintaining the data needed, and 

completing and reviewing the 

information collection. Send comments 

regarding this burden estimate or any 

other aspect of these information 

collections, including suggestions for 

reducing the burden, to the Records and 

FOIA/Privacy Services Branch (T–5 

F52), U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, or by Internet electronic mail to 

infocollects@nrc.gov; and to the Desk 

Officer, Office of Information and 

Regulatory Affairs, NEOB–10202, 

(3150–0011), Office of Management and 

Budget, Washington, DC 20503. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a request for information or an 

information collection requirement 

unless the requesting document 

displays a currently valid OMB control 

number. 

Sample Technical Specifications 

Steam Generator (SG) Program 

A SG Program shall be established 

and implemented to ensure that SG tube 

integrity is maintained. The SG Program 

shall include the following provisions:

a. Provisions for condition-monitoring 

assessments. A condition-monitoring 

assessment is an evaluation of the ‘‘asfound’’ condition of the tubing with 

respect to the performance criteria for 

structural integrity and accidentinduced leakage. The ‘‘as-found’’

condition refers to the condition of the 

tubing during a SG inspection outage, as 

determined from the inservice 

inspection results or by other means, 

prior to the plugging of tubes. 

Condition-monitoring assessments shall 

be conducted during each outage during 

which the SG tubes are inspected or 

plugged to confirm that the performance 

criteria are being met. 

b. Performance criteria for SG tube 

integrity. SG tube integrity shall be 

maintained by meeting the performance 

criteria for tube structural integrity, 

accident-induced leakage, and 

operational LEAKAGE. 

1. Structural integrity performance 

criterion: All inservice SG tubes shall 

retain structural integrity over the full 

range of normal operating conditions 

(including startup, operation in the 

power range, hot standby, and 

cooldown and all anticipated transients 

included in the design specifications) 

and design-basis accidents. This 

includes retaining a safety factor of 3.0 

against burst under the normal steady 

state full-power operation primary-tosecondary pressure differential and a 

safety factor of 1.4 against burst applied 

to the design-basis accident primary-tosecondary pressure differentials. Apart 

from the above requirements, additional 

loading conditions associated with the 

design-basis accidents, or combination 

of accidents in accordance with the 

60196 Federal Register / Vol. 69, No. 194 / Thursday, October 7, 2004 / Notices 

design and licensing basis, shall also be 

evaluated to determine if the associated 

loads contribute significantly to burst or 

collapse. In the assessment of tube 

integrity, those loads that do 

significantly affect burst or collapse 

shall be determined and assessed in 

combination with the loads due to 

pressure with a safety factor of 1.2 on 

the combined primary loads and 1.0 on 

axial secondary loads. 

2. Accident-induced leakage 

performance criterion: The primary-tosecondary accident-induced leakage rate 

for any design-basis accident, other than 

a SG tube rupture, shall not exceed the 

rates assumed in the accident analysis 

for total leakage rate from all SGs and 

leakage rate from an individual SG. 

Accident-induced leakage is not to 

exceed [licensee to insert value] gallons 

per day through each SG and [licensee 

to insert value] gallons per day through 

all SGs. 

3. The operational LEAKAGE 

performance criterion is specified in 

limiting condition for operation (LCO) 

[licensee to insert reference to 

appropriate LCO. For limits currently 

greater than 150 gallons per day, the 

LCO limit should be lowered to a value 

less than or equal to 150 gallons per 

day.] 

c. SG tube repair criteria. Tubes found 

by inservice inspection to contain flaws 

with a depth equal to or exceeding 40 

percent of the nominal tube wall 

thickness shall be plugged. 

d. Provisions for SG tube inspections. 

Periodic SG tube inspections shall be 

performed. The number and portions of 

the tubes inspected and the method of 

inspection shall be performed with the 

objective of detecting flaws of any type 

(for example, volumetric flaws, axial 

and circumferential cracks) that may be 

present along the length of the tube, 

from the tube-to-tubesheet weld at the 

tube inlet to the tube-to-tubesheet weld 

at the tube outlet, and that may satisfy 

the applicable tube repair criteria. The 

tube-to-tubesheet weld is not part of the 

tube. In addition to meeting 

requirements d.1, d.2, and d.3 below, 

the inspection scope, inspection 

methods, and inspection intervals shall 

be such as to ensure that SG tube 

integrity is maintained until the next SG 

inspection. An assessment of 

degradation shall be performed to 

determine the type and location of flaws 

to which the tubes may be susceptible 

and, based on this assessment, to 

determine which inspection methods 

need to be employed and at what 

locations. 

1. Inspect 100 percent of the tubes in 

each SG during the first refueling outage 

following SG replacement. 

2. Inspect 100 percent of the tubes at 

sequential periods of [for licensees with 

thermally treated Alloy 690 tubes, insert 

‘‘144, 108, 72, and thereafter 60 effective 

full-power months;’’ for licensees with 

thermally treated Alloy 600 tubes, insert 

‘‘120, 90, and thereafter 60 effective fullpower months;’’ for licensees with millannealed Alloy 600 tubes, insert ‘‘60 

effective full-power months;’’]. The first 

sequential period shall be considered to 

begin after the first inservice inspection 

of the SGs. In addition, inspect 50 

percent of the tubes by the refueling 

outage nearest the midpoint of the 

period and the remaining 50 percent by 

the refueling outage nearest the end of 

the period. No SG shall operate for more 

than [for licensees with thermally 

treated Alloy 690 tubes, insert ‘‘72 

effective full-power months or three 

refueling outages;’’ for licensees with 

thermally treated Alloy 600 tubes, insert 

‘‘48 effective full-power months or two 

refueling outages;’’ for licensees with 

mill-annealed Alloy 600 tubes, insert 

‘‘24 effective full-power months or each 

refueling outage’’ (whichever is less)] 

without being inspected.

3. If crack indications are found in 

any SG tube, then the next inspection 

for each SG for the degradation 

mechanism that caused the crack 

indication shall not exceed 24 effective 

full-power months or one refueling 

outage (whichever is less). If definitive 

information, such as from examination 

of a pulled tube, diagnostic 

nondestructive testing, or an 

engineering evaluation indicates that a 

cracklike indication is not associated 

with a crack or cracks, then the 

indication need not be treated as a 

crack. 

e. Provisions for monitoring 

operational primary-to-secondary 

leakage. 

Steam Generator (SG) Tube Inspection 

Report 

A report shall be submitted within 

180 days of the initial entry into MODE 

4 following completion of the 

inspection. The report shall include: 

a. The scope of inspection performed 

on each SG. 

b. Active degradation mechanisms 

found. 

c. Nondestructive examination 

techniques utilized for each degradation 

mechanism. 

d. Location, orientation (if linear), and 

measured sizes (if available) of serviceinduced indications. 

e. Number of tubes plugged during the 

inspection outage for each active 

degradation mechanism. 

f. Total number and percentage of 

tubes plugged to date. 

g. The results of condition 

monitoring, including the results of tube 

pulls and in-situ testing. 

End 

Documents may be examined, and/or 

copied for a fee, at the NRC’s Public 

Document Room at One White Flint 

North, 11555 Rockville Pike (first floor), 

Rockville, Maryland. Publicly available 

records will be accessible electronically 

from the Agencywide Documents 

Access and Management System 

(ADAMS) Public Electronic Reading 

Room on the Internet at the NRC Web 

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

index.html. If you do not have access to 

ADAMS or if you have problems in 

accessing the documents in ADAMS, 

contact the NRC Public Document Room 

(PDR) reference staff at 1–800–397–4209 

or 301–415–4737 or by e-mail to 

pdr@nrc.gov.

Dated in Rockville, Maryland, this 30 day 

of September, 2004.

For the Nuclear Regulatory Commission. 

Francis M. Costello, 

Acting Branch Chief, Reactor Operations 

Branch, Division of Inspection Program 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 04–22546 Filed 10–6–04; 8:45 am] 

BILLING CODE 7590–01–P

OVERSEAS PRIVATE INVESTMENT 

CORPORATION 

November 4, 2004, Public Hearing; 

Sunshine Act

TIME AND DATE: 2 p.m., Thursday, 

November 4, 2004.

PLACE: Offices of the Corporation, 

Twelfth Floor Board Room, 1100 New 

York Avenue, NW., Washington, DC.

STATUS: Hearing open to the public at 2 

p.m.

PURPOSE: Public Hearing in conjunction 

with each meeting of OPIC’s Board of 

Directors, to afford an opportunity for 

any person to present views regarding 

the activities of the Corporation.

PROCEDURES: Individuals wishing to 

address the hearing orally must provide 

advance notice to OPIC’s Corporate 

Secretary no later than 5 p.m., Monday, 

October 25, 2004. The notice must 

include the individual’s name, title, 

organization, address, and telephone 

number, and a concise summary of the 

subject matter to be presented. 

Oral presentations may not exceed ten 

(10) minutes. The time for individual 

presentations may be reduced 

proportionately, if necessary, to afford 

all participants who have submitted a