ML042260398
| ML042260398 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/13/2004 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GL-98-005 | |
| Download: ML042260398 (5) | |
Text
August 13, 2004 10 CFR 50.55a(a)(3)(i)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of ) Docket No. 50-259 Tennessee Valley Authority )
BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 1 - AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI AND AUGMENTED INSPECTIONS - RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REQUEST FOR RELIEF, 1-ISI-19, REGARDING REACTOR PRESSURE VESSEL (RPV) CIRCUMFERENTIAL SHELL WELDS In recent discussions with the NRC, the NRC Staff requested that TVA provide additional information to support its review of BFN Unit 1 Relief Request 1-ISI-19. BFN Unit 1 Relief Request 1-ISI-19 requests relief from the provisions of ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item B1.11, requiring the examination of reactor vessel circumferential shell welds, and the (expedited) augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A) for vessel circumferential welds. The additional information requested is provided in Enclosure 1 of this letter.
TVA submitted Relief Request 1-ISI-19 by letter dated May 12, 2004 (Reference 1). That request was based on the guidance contained in NRC Generic Letter (GL) 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds. Relief Request 1-ISI-19 was also consistent with TVAs corresponding request for BFN Unit 2, submitted by letter dated March 24, 2000 (Reference 2) and approved by the NRC by letter dated August 14, 2000 (Reference 3).
U.S. Nuclear Regulatory Commission Page 2 August 13, 2004 However, based on their review, the NRC Staff indicated it needed additional information to complete its review of BFN Unit 1 Relief Request 1-ISI-19. The requested information is provided in the enclosure to this letter.
There are no new commitments contained in this letter. If you have any questions, please telephone me at (256) 729-2636.
Sincerely, Original signed by:
T. E. Abney Manager of Licensing and Industry Affairs
References:
- 1. TVA letter, T. E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) - Unit 1 - American Society Of Mechanical Engineers (ASME)Section XI And Augmented Inspections -
Request For Relief, 1-ISI-19, Regarding Reactor Pressure Vessel (RPV) Circumferential Shell Welds, dated May 12, 2004.
- 2. TVA letter, T. E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) - Unit 2 - American Society Of Mechanical Engineers (ASME)Section XI and Augmented Inspections -
Request for Relief, 2-ISI-9, Regarding Reactor Pressure Vessel (RPV) Circumferential Shell Welds, (TAC No.
MA8424), dated March 24, 2000.
- 3. NRC letter, R. P. Correia to J. A. Scalice (TVA), Browns Ferry Nuclear Plant Unit 2, Relief Request 2-ISI-9, Alternatives for Examination of Reactor Pressure Vessel Shell Welds (TAC No. MA8424), dated August 14, 2000.
cc: See Page 3
U.S. Nuclear Regulatory Commission Page 3 August 13, 2004 Enclosure cc: (Enclosure):
(Via NRC Electronic Distribution)
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Stephen J. Cahill, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 1 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
SECTION XI, INSERVICE (ISI) AND AUGMENTED INSPECTION PROGRAM (FIRST TEN YEAR INSPECTION INTERVAL)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST FOR RELIEF 1-ISI-19 The following provides TVAs responses to the NRC Staff questions concerning BFN Unit 1 Relief Request 1-ISI-19 submitted to the NRC by letter dated May 12, 2004 (Reference 1).
(1) Did the licensee perform a volumetric examination during the first ten year inspection interval of all shell welds in accordance with the requirements of paragraph IWA-2500, of ASME Section XI, 1995 Edition, 1996 Addenda?
Response
No. The shell weld examinations performed to date were performed prior to the current BFN Unit 1 extended outage which began in 1985. These examinations were performed in accordance with the initial code of record, ASME Section XI, 1971 Edition, Summer 1971 Addenda and later updated for the second period to the ASME Section XI 1974 Edition with Addenda through Summer 1975, with no indications identified.
(2) Augmented volumetric examination of RPV shell welds is required per 10 CFR 50.55a (g)(6)(ii)(A). The licensee should identify, if this examination has been performed or when this examination will be performed for all axial shell welds.
Response
The reactor vessel shell weld examinations required in accordance with 10 CFR 50.55a(g)(6)(ii)(A) will be performed for all axial shell welds prior to restart of BFN Unit 1 from its current outage.
(3) Table 2 lists the 32 effective full power years fluence value for weld C-1-2. Also, the submittal stated that the fluence value was obtained using a methodology consistent with the guidance in RG 1.190. Please substantiate the proposed fluence value and the method used by providing a reference or statement
E-2 for the method used and the associated parameter values.
Response
General Electric Nuclear Energy (GENE) performed the fluence calculations for BFN Unit 1 using the 2-D discrete ordinates methodology described in NEDO-32983-A, "]GE Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations, Revision 0, December 2001. The NRC reviewed and approved that methodology, transmitting its safety evaluation by letter from S. A. Richards (NRC) to J. F. Klapproth (GE) by letter dated September 14, 2001. The NRC acknowledged that GEs method adheres to the guidance contained in NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.
Key inputs to the calculation and results were as follows:
KEY INPUTS Number of fuel assemblies in the core 764 Fuel assembly type GE14 Assumed operation 32 EFPY Vessel thickness 6.13 inches KEY RESULTS RPV ID peak flux (>1.0Mev) 1.40 E9 n/ cm2-sec RPV ID peak flux azimuth 27.25° RPV ID peak flux location 100.9 inches above bottom of active fuel Ratio peak/Girth Weld location 0.81 RPV ID peak surface fluence1 1.42 E18 n/cm2 RPV ID peak surface fluence at Girth Weld location 1.15 E18 n/cm2 Girth Weld peak fluence at 1/4T
.0799E19 n/cm2 Notes:
- 1. RPV ID peak surface fluence raised 1% (1.41E18 n/cm2 to 1.42 E18 n/cm2) to account for MELLLA+ operation.
E-3
Reference:
- 1.
TVA letter, T. E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) - Unit 1 - American Society Of Mechanical Engineers (ASME)
Section XI And Augmented Inspections - Request For Relief, 1-ISI-19, Regarding Reactor Pressure Vessel (RPV) Circumferential Shell Welds, dated May 12, 2004.