ML042600522

Summary of Telephone Conference Held on August 19, 2004, Between the U.S. Nuclear Regulatory Commission and the Tennessee Valley Authority Concerning Draft Requests for Additional Information on Browns Ferry Nuclear Plant, Units 1, 2 & 3 LR

ML042600522
Person / Time
Site:	Browns Ferry
Issue date:	09/16/2004
From:	Yoira Diaz-Sanabria NRC/NRR/DRIP/RLEP
To:	Office of Nuclear Reactor Regulation
DIAZ-SANABRIA, Y /NRR/DRIP/RLEP 415-1594
References
TAC MC1704, TAC MC1705, TAC MC1706
Download: ML042600522 (11)

Text

September 16, 2004 LICENSEE: Tennessee Valley Authority FACILITY: Browns Ferry Nuclear Plant Units 1, 2 and 3

SUMMARY

OF TELEPHONE CONFERENCE HELD ON AUGUST 19, 2004, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND THE TENNESSEE VALLEY AUTHORITY CONCERNING DRAFT REQUESTS FOR ADDITIONAL INFORMATION ON BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 LICENSE RENEWAL APPLICATION (TAC NOS. MC1704, MC1705 AND MC1706)

The U.S. Nuclear Regulatory Commission staff and representatives of Tennessee Valley Authority (TVA or the applicant) held a telephone conference on August 19, 2004, to discuss the draft requests for additional information (D-RAIs) related to Section 3.1.2 of the Browns Ferry Nuclear Plant (BFN) license renewal application.

The conference call was useful in clarifying the intent of the staffs questions. On the basis of the discussion, the applicant was able to understand the staff's questions. No staff decisions were made during the telephone conference. In some cases, the applicant agreed to provide information for clarification. contains a listing of the D-RAIs discussed with the applicant, including a description on the status of the items. Enclosure 2 provides a list of the telephone conference participants. The applicant has had an opportunity to comment on this summary.

/RA/

Yoira K. Diaz Sanabria, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket Nos.: 50-259, 50-260 and 50-296

Enclosures:

As stated cc w/encls: See next page

September 16, 2004 LICENSEE: Tennessee Valley Authority FACILITY: Browns Ferry Nuclear Plant Units 1, 2 and 3

SUMMARY

OF TELEPHONE CONFERENCE HELD ON AUGUST 19, 2004, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND THE TENNESSEE VALLEY AUTHORITY CONCERNING DRAFT REQUESTS FOR ADDITIONAL INFORMATION ON BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 LICENSE RENEWAL APPLICATION (TAC NOS. MC1704, MC1705 AND MC1706)

The U.S. Nuclear Regulatory Commission staff and representatives of Tennessee Valley Authority (TVA or the applicant) held a telephone conference on August 19, 2004, to discuss the draft requests for additional information (D-RAIs) related to Section 3.1.2 of the Browns Ferry Nuclear Plant (BFN) license renewal application.

The conference call was useful in clarifying the intent of the staffs questions. On the basis of the discussion, the applicant was able to understand the staff's questions. No staff decisions were made during the telephone conference. In some cases, the applicant agreed to provide information for clarification. contains a listing of the D-RAIs discussed with the applicant, including a description on the status of the items. Enclosure 2 provides a list of the telephone conference participants. The applicant has had an opportunity to comment on this summary.

/RA/

Yoira K. Diaz Sanabria, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket Nos.: 50-259, 50-260 and 50-296

Enclosures:

As stated cc w/encls: See next page Accession No.: ML042600522 DISTRIBUTION:

See next page Document Name:C:\ORPCheckout\FileNET\ML042600522.wpd OFFICE PM:RLEP LA:RLEP SC:RLEP NAME Y. Diaz-Sanabria M. Jenkins S. Lee DATE 9/15/04 9/15/04 9/16/04 OFFICIAL RECORD COPY

DISTRIBUTION: Dated: September 16, 2004 Accession No: ML042600522 HARD COPY RLEP RF Yoira Diaz Sanabria (PM)

E-MAIL:

RidsNrrDSSA RidsNrrDe RLEP staff S. Black Y.L. (Renee) Li C. Li K. Jabbour A. Hodgdon R. Weisman C. Carpenter A. Howe R. Subbaratnam G. Cranston G. Cheruvenki B. Elliot

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 LICENSE RENEWAL APPLICATION DRAFT REQUEST FOR ADDITIONAL INFORMATION (D-RAI)

AGING MANAGEMENT OF REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SECTIONS 3.1.2 Reactor Vessel Internals Core Shroud and Core Plate D-RAI-3.1.2.1-6 Table 3.1.2.1 (a) According to Section IV-B.1.1-d/B1.1.4 of NUREG-1801, augmented inspection of access hole covers is required for Alloy 600 materials and Alloy 182 welds. The applicant should provide details on the type and frequency of inspection of the access hole covers.

(b) According to Section IV-B.1.1-d of NUREG-1801, irradiation assisted stress corrosion cracking is an aging effect for core shroud components. Please provide an explanation for excluding this aging mechanism in Table 3.1.2.1. Provide details of the AMP that will be implemented on this system.

(c) Describe plant-specific experience related to IGSCC cracking of the stainless steel and Inconel components in the core shroud, and shroud support access hole covers. Provide details on any occurrence of IGSCC cracking, and the effective AMP that will be implemented on these systems at BFN Units.

(d) The applicant should address the plant-specific experience on sudden increase in RCS water conductivity due to a leak in condensate and or reactor water clean up systems. Provide information on the impact of sudden increase in RCS water conductivity on IGSCC of core shroud welds.

(e) Provide information on verification methods to monitor the effectiveness of the hydrogen water chemistry program. Explain the methodology of ensuring hydrogen availability in the core shroud region. If ECP probes are not used to monitor availability of hydrogen, explain the validity of using secondary parameters (i.e., main steam/feedwater oxygen levels) to assess the hydrogen availability at core shroud welds.

Discussion: The applicant indicated that the question is clear. This question will be sent as RAI 3.1.2.1-6.

Enclosure 1

Core Spray Spargers and piping D-RAI-3.1.2.2-7 Table 3.1.2.2 (a) According to the Section IV-B.1.3-a of NUREG-1801, irradiation assisted stress corrosion cracking is an aging effect for the core spray lines and spargers components. Please provide an explanation for excluding this aging effect in Table 3.1.2.1. Provide details of the AMP that will be implemented on this system.

(b) Core Spray piping and spargers contain crevice conditions in some weld areas. Explain the methodology of ensuring hydrogen availability in these systems. Since the presence of crevice conditions enhances the occurrence of IGSCC, the applicant should provide details on the type and extent of inspections to identify IGSCC, and the mitigation techniques at BFN Units 2 and 3.

Discussion: The applicant indicated that the question is clear. This question will be sent as RAI 3.1.2.2-7.

Reactor Vessel Internals Dry Tube and Guide Tube D-RAI-3.1.2.2-8 Table 3.1.2.2 (a) According to Section IV-B.1.6-a of NUREG-1801, irradiation assisted stress corrosion cracking is an aging effect for dry tube and guide tube components. Please provide an explanation for excluding this aging effect in Table 3.1.2.1. Provide information on the AMP that will be implemented on this system.

(b) The AMP for the dry tube and guide tube components addressed in the application references BWRVIP-47. Table 3.1-2 of BWRVIP-47 indicates that some of the incore housing guide tubes and dry tubes for BFN Units 2 and 3 experienced cracking and were subsequently replaced with materials resistant to cracking. Provide information on the type and grade of the replaced material, and its performance at BFN Units 2 and 3. The staff requests additional information on the type and extent of subsequent inspections of the dry tubes and guide tubes for BFN Units 2 and 3. The applicant should also address any existence of cracks in BFN Unit 1 dry tubes and guide tubes. Identify the method and frequency of inspection of the BFN Unit 1 dry tubes and guide tubes. Provide information on any mitigating techniques that will be implemented to reduce the susceptibility to cracking in Unit 1 dry tubes and guide tubes.

(c) According to Section 2.2.1.2 of BWRVIP-47, furnace-sensitized stainless steel stub tubes are more susceptible to IGSCC. The applicant should provide information on any existing furnace-sensitized stub tubes at BFN Units. Provide details on the AMP that will be implemented for the furnace-sensitized stainless steel stub tubes at the BFN Units. Identify whether any furnace-sensitized stainless steel stub tubes have previously experienced cracking due to SCC, IGSCC or cyclic loading, and the extent of the cracking. Identify the method and frequency of inspection.

(d) According to the Section 2.2.1.2 of BWRVIP-47, weld metal 182 is more susceptible to IGSCC. Provide details on the AMP for components that have 182 weld metal in these systems at BFN Units. Identify whether any 182 weld metals have previously experienced cracking due to SCC, IGSCC or cyclic loading, and the extent of cracking. Identify the method and frequency of inspection.

Discussion: The applicant indicated that the question is clear. This question will be sent as RAI 3.1.2.2-8.

Jet Pump Assembly D-RAI-3.1.2.2-9 Table 3.1.2.2 Provide information on any existing Cast Austenitic Stainless Steel (CASS) jet pump components. The applicant should provide the information on the jet pump components:

(a) Information on type of casting (i.e; centrifugal or static)

(b) The composition of CASS (i.e; Molybdenum content and delta ferrite values)

(c) Previous plant-specific experience regarding the cracked components and type and extent of subsequent inspection of CASS jet pump components due to neutron and thermal embrittlement. The fluence values should be based on the end of the extended period of operation and power uprate.

(d) The LRA should address any technical specification changes related to jet pump components.

Discussion: The applicant indicated that the question is clear. This question will be sent as RAI 3.1.2.2-9.

Top Guide D-RAI-3.1.2.2-10 Table 3.1.2.2 According to Section IV -B.1.2-a of NUREG-1801, irradiation assisted stress corrosion cracking is an aging effect for top guide components. Please provide explanation for excluding this issue in Table 3.1.2.1. Provide details on the AMP that will be implemented on this system.

Discussion: The applicant indicated that the question is clear. This question will be sent as RAI 3.1.2.2-10.

Reactor Pressure Vessel (RPV) Vents and Drains Piping D-RAI 3.1.2.3-2 The LRA should identify aging effects for the carbon steel drain line penetrations exposed to reactor coolant water up to 288°C (550° F). Such drain lines are likely to experience loss of material due to corrosion. This assessment is consistent with Item D2.1-a, Chapter V.D2 of NUREG-1801. Explain why loss of material due to corrosion is not considered as an aging effect for these components, or provide a program for managing such effect.

Table 3.1.2.3 -RPV Vents and Drains, identifies a one-time inspection specified in B.2.1.29 as a part of the AMP. According to B.2.1.29 a one time inspection is applicable for piping and fittings with diameter less than 4 inches Nominal Pipe Size (NPS). Identify whether the reactor vents and drain systems have previously experienced cracking due to SCC, IGSCC or cyclic loading, and the extent of cracking. Identify the method of the one-time inspection. The LRA does not address the type and the frequency of the inspection requirements as a part of the AMP, for piping greater than 4 inches NPS.

Discussion: The applicant indicated that the question is clear. This question will be sent as RAI 3.1.2.3-2.

Reactor Recirculation System Piping D-RAI 3.1.2.4-2 Table 3.1.2.4 identifies a one-time inspection specified in B.2.1.29 as a part of the AMP.

According to B.2.1.29 one-time inspection is applicable for piping and fittings with diameter less than 4 inches NPS. Identify whether the reactor recirculation system has previously experienced cracking due to SCC, IGSCC or cyclic loading, and the extent of cracking. Identify the method of the one time inspection. The LRA does not address the type and frequency of the inspection requirements as a part of the AMP, for piping greater than 4 inches NPS.

Discussion: The applicant indicated that the question is clear. This question will be sent as RAI 3.1.2.4-2.

LIST OF PARTICIPANTS FOR TELEPHONE CONFERENCE ON DRAFT REQUESTS FOR ADDITIONAL INFORMATION August 19, 2004 Participants Affiliation Yoira Diaz U.S. Nuclear Regulatory Commission (NRC)

Barry Elliot NRC Ganesh Cheruvenki NRC Ram Subbaratnam NRC Gregory Cranston NRC Ken Brune Tennessee Valley Authority (TVA)

Valarie Smith TVA Terry Knuettel TVA Mickey Hamby TVA Enclosure 2

BROWNS FERRY NUCLEAR PLANT Tennessee Valley Authority cc:

Mr. Karl W. Singer, Senior Vice President Mr. Robert G. Jones Nuclear Operations Browns Ferry Unit 1 Plant Restart Manager Tennessee Valley Authority Browns Ferry Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Decatur, AL 35609 Mr. James E. Maddox, Vice President Mr. Mark J. Burzynski, Manager Engineering & Technical Services Nuclear Licensing Tennessee Valley Authority Tennessee Valley Authority 6A Lookout Place 4X Blue Ridge 1101 Market Street 1101 Market Street Chattanooga, TN 37402-2801 Chattanooga, TN 37402-2801 Mr. Ashok S. Bhatnagar, Site Vice President Mr. Timothy E. Abney, Manager Browns Ferry Nuclear Plant Licensing and Industry Affairs Tennessee Valley Authority Browns Ferry Nuclear Plant P.O. Box 2000 Tennessee Valley Authority Decatur, AL 35609 P.O. Box 2000 Decatur, AL 35609 General Counsel Tennessee Valley Authority Mr. Bobby L. Holbrook ET 11A Senior Resident Inspector 400 West Summit Hill Drive U.S. Nuclear Regulatory Commission Knoxville, TN 37902 Browns Ferry Nuclear Plant 10833 Shaw Road Mr. Thomas J. Niessen, Acting General Athens, AL 35611 Manager Nuclear Assurance State Health Officer Tennessee Valley Authority Alabama Dept. of Public Health 6A Lookout Place RSA Tower - Administration 1101 Market Street Suite 1552 Chattanooga, TN 37402-2801 P.O. Box 303017 Montgomery, AL 36130-3017 Mr. Michael D. Skaggs Plant Manager Browns Ferry Nuclear Plant Chairman Tennessee Valley Authority Limestone County Commission P.O. Box 2000 310 West Washington Street Decatur, AL 35609 Athens, AL 35611 Mr. Jon R. Rupert, Vice President Mr. Fred Emerson Browns Ferry Unit 1 Restart Nuclear Energy Institute Browns Ferry Nuclear Plant 1776 I St., NW, Suite 400 Tennessee Valley Authority Washington, DC 20006-2708 P.O. Box 2000 Decatur, AL 35609