ML041480033
| ML041480033 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/26/2004 |
| From: | Stephanie Coffin NRC/NRR/DLPM/LPD2 |
| To: | Stinson L Southern Nuclear Operating Co |
| Peters S, NRR/DLPM, 415-1842 | |
| References | |
| TAC MC1598, TAC MC1599 | |
| Download: ML041480033 (8) | |
Text
May 26, 2004 Mr. L. M. Stinson Vice President - Farley Project Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham, Alabama 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 RE: REQUEST FOR RELIEF FOR USE OF ALTERNATE CODE CASE FOR NEW CONSTRUCTION OF REPLACEMENT REACTOR VESSEL HEAD ADAPTERS (TAC NOS.
Dear Mr. Stinson:
By letter dated December 19, 2003, as supplemented by letter dated March 5, 2004, Southern Nuclear Operating Company, Inc. (the licensee), pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(a)(3)(i), submitted a relief request from the requirements of Section III of the 1998 edition and 2000 addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code). Specifically, the licensee requested to use design stress intensity values included in Code Case N-698 for the manufacture of reactor pressure vessel head adapters at Farley Nuclear Plant (FNP), Units 1 and 2.
The NRC staff has reviewed the information provided for this relief request and finds the proposed use of the design stress intensity values, as stated in Code Case N-698, acceptable.
The NRC staffs Safety Evaluation is provided in the Enclosure. Pursuant to 10 CFR 50.55a(a)(3)(i), the NRC grants relief for FNP, Units 1 and 2 for the third 10-year inservice inspection interval on the basis that the proposed alternative provides an acceptable level of quality and safety. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Sincerely,
/RA/
Stephanie M. Coffin, Acting Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosure:
As stated cc w/encl: See next page
May 26, 2004 Mr. L. M. Stinson Vice President - Farley Project Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham, Alabama 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 RE: REQUEST FOR RELIEF FOR USE OF ALTERNATE CODE CASE FOR NEW CONSTRUCTION OF REPLACEMENT REACTOR VESSEL HEAD ADAPTERS (TAC NOS.
Dear Mr. Stinson:
By letter dated December 19, 2003, as supplemented by letter dated March 5, 2004, Southern Nuclear Operating Company, Inc. (the licensee), pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(a)(3)(i), submitted a relief request from the requirements of Section III of the 1998 edition and 2000 addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code). Specifically, the licensee requested to use design stress intensity values included in Code Case N-698 for the manufacture of reactor pressure vessel head adapters at Farley Nuclear Plant (FNP), Units 1 and 2.
The NRC staff has reviewed the information provided for this relief request and finds the proposed use of the design stress intensity values, as stated in Code Case N-698, acceptable.
The NRC staffs Safety Evaluation is provided in the Enclosure. Pursuant to 10 CFR 50.55a(a)(3)(i), the NRC grants relief for FNP, Units 1 and 2 for the third 10-year inservice inspection interval on the basis that the proposed alternative provides an acceptable level of quality and safety. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Sincerely,
/RA/
Stephanie M. Coffin, Acting Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosure:
As stated DISTRIBUTION: See next page cc w/encl: See next page
- See Previous Concurrence Adams Accession No.: ML041480033
- No Major Changes to SE OFFICE PDII-1/PM PDII-1/LA (A)
EMCB/SC OGC PDII-1/SC (A)
NAME SPeters**
DClarke**
TChan*
JHull**
SCoffin DATE 05/11/04 05/13/04 03/30/04 05/25/04 5/26/04 OFFICIAL RECORD COPY
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF FOR USE OF ALTERNATE CODE CASE FOR NEW CONSTRUCTION OF REPLACEMENT REACTOR VESSEL HEAD ADAPTERS SOUTHERN NUCLEAR OPERATING COMPANY, INC., ET AL.
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By letter dated December 19, 2003, as supplemented by letter dated March 5, 2004, Southern Nuclear Operating Company, Inc. (SNC, the licensee), pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(a)(3)(i), submitted a request for relief from the requirements of Section III of the 1998 edition and 2000 addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code). Specifically, the licensee requested to use design stress intensity values included in Code Case N-698, Design Stress Intensities and Yield Strength Values for UNS N06690 with a Minimum Specified Yield Strength of 35 ksi, Class 1 ComponentsSection III, Division 1, for the manufacture of reactor pressure vessel (RPV) head adapters at Farley Nuclear Plant (FNP), Units 1 and 2.
2.0 REGULATORY EVALUATION
2.1 Applicable Requirements As stated in 10 CFR 50.55a, it is required that systems and components of nuclear power plants are designed and constructed in accordance with the rules provided in the ASME Code Section III, Division 1. Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the ASME Code requirements can be used when authorized by the NRC if the licensee demonstrates that:
(i) the proposed alternative would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.2 System/Component(s) for Which Relief is Requested:
Relief is requested for RPV closure heads manufactured to the requirements of Section III of 1998 edition and 2000 addenda of the ASME Boiler and Pressure Vessel Code.
Enclosure 2.3 Code Requirements for which Relief is Requested:
The replacement reactor vessel head manufactured to the requirements of Section III of the 1998 edition and 2000 addenda of the ASME Code, Section NB-3112.4 requires the use of design stress intensity values for materials listed in the tables of Section II, Part D, Subpart 1.
Table 2B of Section II, Part D contains the design stress intensity values for Type Designation UNS N06690 for the SB-167 specification for Class 1 components.
2.4 Licensees Proposed Alternative:
Section II, Part B, Table 3 of the SB-167 specification provides an UNS N06690 yield strength that is dependent on manufactured condition. For seamless pipe and tube that are less than or equal to 5.0 inch outside diameter, Table 3 specifies a minimum yield strength of 30 kilopounds per square inch (ksi) for hot-worked or hot-worked annealed and 35 ksi for cold-worked annealed.
The UNS N06690 adapters in the replacement head are manufactured using a hot-working process. Evaluation of the hot-worked UNS N06690 tubing has shown that a minimum yield strength of 35 ksi is inherent to the material. Because of the large margin in the minimum yield strength in ASME Section II, Table 3, an ASME Code Case was needed to permit the use of higher design stress intensity values (23.3 ksi in lieu of 20 ksi) when the minimum yield strength is specified as 35 ksi for hot-worked UNS N06690 Class 1 components.
Therefore, ASME developed and approved ASME Code Case N-698 Design Stress Intensities and Yield Strength Values for UNS N06690 with a Minimum Specified Yield Strength of 35 ksi, Class 1 ComponentsSection III, Division 1, permitting the use of higher stress intensities associated with 35 ksi minimum yield strength, hot-worked UNS N06690 SB-167. The licensee is requesting that the design stress intensity values of Code Case N-698 be approved pursuant to 10 CFR 50.55a(a)(3)(i).
3.0 TECHNICAL EVALUATION
3.1 Licensees Basis for Relief:
The use of UNS N06690 in the SB-167 specification has been previously approved by the ASME as Code Case N-474-2, Design Stress Intensities and Yield Strength Values for UNS N06690 with a Minimum Specified Yield Strength of 35 ksi, Class I ComponentsSection III, Division 1, and accepted by the NRC in Regulatory Guide (RG) 1.84, Revision 31, dated March 1999. The design stress intensities and yield strength of Code Case N-474-2 were incorporated into the 1999 addenda for the annealed condition without specification of type of anneal.
Code Case N-525, Design Stress Intensities and Yield Strength Values for UNS N06690 with a Minimum Specified Yield Strength of 30 ksi, Class I ComponentsSection III, Division I was accepted by the NRC staff for use in the construction of components in RG 1.84, Revision 32, dated June 2003, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III," which is incorporated into 10 CFR 50.55a(b) by reference.
However, Code Case N-525 does not address the manufactured condition of the UNS N06690 for which a 30 ksi yield strength is specified. Table 3 of the SB-167 specifications in Section II, Part B implies that the specified 30 ksi yield strength in Code Case N-525 applies to hot-worked UNS N06690. Similarly, it implies the specified 35 ksi yield strength of Code Case N-474-2 and the 1999 addenda would apply to cold-worked UNS N06690.
For the hot-worked UNS N06690, Code Case N-698 adjusts the yield strength values to 35-ksi minimum and provides associated design stress intensity values that coincide with those of Section II, Part D, Subpart 1 of the 1999 addenda.
The use of Code Case N-698 is considered conservative in that associated design analyses will be performed utilizing the most recent determination of material strength values.
Implementation of Code Case N-698 would therefore represent a suitable alternate to Section II, Part D for UNS N06690 SB-167 material and provide an acceptable level of quality and safety to assure pressure boundary requirements are maintained. Therefore, approval should be granted pursuant to 10 CFR 50.55a(a)(3)(i).
3.2 NRC Staff Evaluation Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested relief from ASME Section III requirements for the manufacture of RPV closure heads at FNP, Units 1 and 2. As an alternative, the licensee proposed the use of the design stress intensity values of Code Case N-698. Specifically, SNC proposed using the design stress intensity value of 23.3 ksi for all temperatures of UNS N06690 tubing.
The ASME code requires replacement RPV heads to be manufactured to the requirements of Section III of the 1998 edition and 2000 addenda of the Code. Section NB-3112.4 requires the use of the design stress intensity values for construction materials listed in the tables of Section II, Part D, Subpart 1. For material type UNS N06690 of the SB-167 specification, the yield strength is dependent on the manufactured condition.Section II, Part B, Table 3 specifies a minimum yield strength of 30 ksi for hot-working or hot-worked annealed and 35 ksi for cold-worked annealed.
The licensee requested the use of UNS N06690 adapters in the replacement head which are manufactured by a hot-working process. Evaluation of hot-worked UNS N06690 tubing has shown that a minimum yield strength of 35-ksi is inherent to this material. Therefore, design stress intensity values of 23.3 ksi as permitted by ASME Code Case N-698 is justified. Further, the ASME Code Case N-474-2, which has been endorsed by the NRC in its RG 1.84 permits the use of design stress intensity values of 23.3 ksi for UNS N06690 materials with a minimum yield strength of 35 ksi. Therefore, the staff finds that proposed use of the design stress intensity value of 23.3 ksi to be acceptable.
4.0 CONCLUSION
Based on the above evaluation, the staff concludes that the proposed alternative discussed in the licensees request for relief provides an acceptable level of quality and safety. Therefore, use of the proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for FNP, Units 1 and 2 for the third 10-year inservice inspection interval. All other ASME Code,Section XI requirements (for which relief was not specifically requested and approved in this relief request) remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: D. Votolato Date: May 26, 2004
Joseph M. Farley Nuclear Plant cc:
Mr. Don E. Grissette General Manager -
Southern Nuclear Operating Company Post Office Box 470 Ashford, Alabama 36312 Mr. B. D. McKinney, Licensing Manager Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201-1295 Mr. M. Stanford Blanton Balch and Bingham Law Firm Post Office Box 306 1710 Sixth Avenue North Birmingham, Alabama 35201 Mr. J. B. Beasley, Jr.
Executive Vice President Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201 State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-1701 Chairman Houston County Commission Post Office Box 6406 Dothan, Alabama 36302 Resident Inspector U.S. Nuclear Regulatory Commission 7388 N. State Highway 95 Columbia, Alabama 36319 William D. Oldfield SAER Supervisor Southern Nuclear Operating Company P. O. Box 470 Ashford, Alabama 36312
DISTRIBUTION FOR JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 RE:
REQUEST FOR RELIEF FOR USE OF ALTERNATE CODE CASE FOR NEW CONSTRUCTION OF REPLACEMENT REACTOR VESSEL HEAD ADAPTERS (TAC NOS.
Dated: May 26, 2004 DISTRIBUTION:
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