1CAN010801, Request for Relief from ASME Section XI, Table IWB-2500-1 Requirements
| ML080250308 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/22/2008 |
| From: | James D Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN010801 | |
| Download: ML080250308 (17) | |
Text
1CAN010801 January 22, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
Request for Relief from ASME Section XI Table IWB-2500-1 Requirements Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51
Dear Sir or Madam:
Pursuant to 10 CFR 50.55a(a)(3), Entergy Operations, Inc. (Entergy) hereby requests approval of proposed alternatives to the requirements of the 1992 Edition of ASME Section XI for Arkansas Nuclear One, Unit 1 (ANO-1). Specifically the requirements of Examination Category B-A, B-F, and B-J of Table IWB-2500-1. Details of the requests are provided in the individual attachments to this letter.
The requests are based upon previously submitted and approved relief requests for other sites, including Beaver Valley and Diablo Canyon.
In accordance with 10 CFR 50.55a(a)(3), the proposed alternatives to the referenced requirements may be approved by the NRC provided an acceptable level of quality and safety are maintained. Entergy believes the proposed alternatives meet this requirement.
The relief requests include one new commitment as summarized in Attachment 4.
Entergy requests approval of the proposed alternatives by October 1, 2008, in order to support the fall 2008 refueling outage. Although this request is neither exigent nor emergency, your prompt review is requested.
Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4619 Dale E. James Manager, Licensing Nuclear Safety Assurance Arkansas Nuclear One
1CAN010801 Page 2 of 2 If you have any questions or require additional information, please contact Bob Clark at 479-858-4663.
Sincerely, DEJ/rwc Attachments:
- 1. Request for Relief ANO1-ISI-009
- 2. Request for Relief ANO1-ISI-010
- 3. Request for Relief ANO1-ISI-011
- 4. List of Regulatory Commitments cc:
Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS O-7 D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health & Human Services P.O. Box 1437 Slot H-30 Little Rock, AR 72203-1437
1CAN010801 Request for Relief ANO1-ISI-009 to 1CAN010801 Page 1 of 4 REQUEST FOR RELIEF ANO1-ISI-009 I.
Code Requirement(s)
The current code of record governing in-service inspection for Arkansas Nuclear One, Unit 1 (ANO-1), is the 1992 Edition. ASME Section XI, 1992 Edition, Table IWB-2500-1, Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, Item B1.30 requires a volumetric examination of essentially 100% of the weld length of the required examination volume as identified in Figure IWB-2500-4. Appendix I of Section XI describes the ultrasonic examination requirements for components.
On September 22, 1999, 10 CFR 50.55a was revised mandating the implementation of the 1995 Edition, through and including the 1996 Addenda of ASME Section XI, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems. These requirements are now implemented, as applicable, in the in-service inspection program for ANO-1.
ASME Section XI, Appendix I, Paragraph I-2110 states in part, Ultrasonic examination procedures, equipment, and personnel used to detect and size flaws in reactor vessels greater than 2 inches in thickness shall be qualified by performance demonstration in accordance with Appendix VIII for the following specific examinations and no other I-2000 requirements apply.
Nozzle and Head Welds excluding Flange Welds Components/Numbers:
01-001 Code Classes:
ASME Code Class 1
References:
ASME Section XI, Division 1, 1992 Edition, Table IWB-2500-1 ASME Section XI, Division 1, 1995 Edition with 1996 Addenda, Appendix I and Appendix VIII NRC Regulatory Guide 1.150, Revision 1 Examination Category:
B-A Item Number(s):
B1.30
==
Description:==
Reactor Vessel Shell to Flange Weld Unit / Inspection Interval Applicability:
Arkansas Nuclear One, Unit 1 / Third (3rd) 10-year interval, 1R21 Refueling Outage to 1CAN010801 Page 2 of 4 ASME Section XI, Appendix I, Paragraph I-2120, entitled Other Vessels, states, Ultrasonic examination of all other vessels greater than 2 inches in thickness shall be conducted in accordance with Article 4 of Section V, as supplemented by Table I-2000-1. This removes the reactor vessel flange-to-shell weld from the more rigorous requirements of Appendix VIII; therefore, it would require examination to the 1992 Edition of ASME Section V, Article 4.
Additionally, Regulatory Guide (RG) 1.150, Revision 1, Ultrasonic Testing of Reactor Vessel Welds during Pre-service and In-service Examinations augments the ASME Section V and XI Code requirements.
II.
Proposed Alternative Examinations In lieu of using the techniques of Section V, Article 4, as supplemented by Section XI, Appendix I and augmented by RG 1.150, Revision 1, when performing volumetric (ultrasonic) examination of the reactor vessel shell-to-flange weld, Entergy proposes using qualified personnel and procedures for remote mechanized examination in accordance with the 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplements 4 and 6, as modified by 10 CFR 50.55a(b)(2)(xv), and demonstrated by the EPRI PDI Program, for the reactor vessel shell-to-flange weld.
These examinations will be conducted with qualified vendor procedures for detection and length/depth sizing of ferritic material with a nominal thickness up to 12.3 inches for single or dual-sided applications.
III. Basis for Alternative The prescriptive, amplitude-based ultrasonic examination techniques of the 1992 Edition of ASME Section V, Article 4, supplemented by Appendix I and augmented by RG 1.150, Revision 1 (hereafter referred to as Article 4), are technically inferior to the performance-based techniques specified in the 1995 Edition with 1996 Addenda of Section XI, Appendix VIII, Supplements 4 and 6, as modified by 10 CFR 50.55a(b)(2)(xv), and demonstrated through the Electric Power Research Institute (EPRI) Performance Demonstration Initiative (PDI) Program (hereafter referred to as Appendix VIII).
The performance-based techniques of Appendix VIII are required for all other Reactor Vessel Shell Weld examinations, having replaced Article 4 techniques.
Radiation exposure will be reduced since the change out of examination devices on the inspection robot will not be necessary to perform the shell-to-flange weld examination.
Additionally, the performance-based techniques of Appendix VIII offer several enhancements over the prescriptive amplitude-based techniques as described below:
(a) Increased Sensitivity to Flaws: The Appendix VIII procedure is more sensitive to flaws because the examination sensitivity level compares to that of an ASME distance amplitude correction (DAC) level of 5 to 10 percent, which is the highest practical level for ultrasonic testing. Examinations in accordance with Article 4 are conducted at a 50 percent DAC for the outer 80 percent of the wall thickness and 20 percent for the inner 20 percent of the wall thickness. The Appendix VIII to 1CAN010801 Page 3 of 4 procedure requires all signals interpreted by the analyst as flaws to be measured and assessed in accordance with the applicable acceptance criteria, regardless of amplitude. This recognizes that some flaws can exhibit a low amplitude response depending on orientation. The Article 4 techniques traditionally have a flaw recording cut-off point of 20 percent DAC.
(b) Demonstrated flaw measurement capability using amplitude-independent sizing techniques: The procedure for the proposed shell-to-flange weld examination has been demonstrated in accordance with Appendix VIII. The proposed procedure complies with ASME Section XI, 1995 Edition with 1996 Addenda, as modified by 10 CFR 50.55a. The procedure has been qualified by time-based sizing techniques such as tip diffraction rather than amplitude-based Article 4 techniques that have been proven inaccurate.
(c) Compatibility of the Appendix VIII examination technique with ANO-1 RPV shell-to-flange geometry and previous ISI examinations for data comparison: The proposed Appendix VIII examination procedure will use the 45 degree beam angle in four orthogonal directions applied to the weld and examination volume by various transducer types, each covering a specified depth range. The incremental scan lines will be 0.5 inches and examination will be conducted to the maximum extent practical.
The previous remote mechanized examination of the shell-to-flange weld was conducted at ANO-1 in 1995. At that time, 45, 60 and 70-degree examination angles were used.
Results were acquired and analyzed using automated ultrasonic examination systems and subsequently evaluated to the applicable criteria in ASME Section XI. Data archival from the previous examination is available for comparison purposes to the proposed examination.
Use of qualified personnel and procedures to the 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplements 4 and 6, as modified by 10 CFR 50.55a(b)(2)(xv) by demonstration through the EPRI PDI Program for the reactor vessel shell-to-flange weld provides an equivalent or better level of quality and safety than the current ASME Section V, Article 4, supplemented by Appendix I and augmented by RG 1.150, Revision 1 requirements.
IV. Conclusion 10 CFR 50.55a(a)(3) states:
Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g) and (h) of this section or portions thereof may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that:
(i)
The proposed alternatives would provide an acceptable level of quality and safety, or (ii)
Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
to 1CAN010801 Page 4 of 4 Use of qualified personnel and procedures to the 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplements 4 and 6, as modified by 10 CFR 50.55a(b)(2)(xv) by demonstration through the EPRI PDI Program for the reactor vessel shell-to-flange weld provides an equivalent or better level of quality and safety than the current ASME Section V, Article 4, supplemented by Appendix I and augmented by RG 1.150, Revision 1 requirements. Therefore, Entergy requests authorization to perform the proposed alternative to the Code requirement pursuant to 10 CFR 50.55a(a)(3) for implementation during the ANO-1, 1R21 refueling outage scheduled for the fall of 2008.
1CAN010801 Request for Relief ANO1-ISI-010 to 1CAN010801 Page 1 of 3 REQUEST FOR RELIEF ANO1-ISI-010 I.
Code Requirement(s)
The current code of record governing in-service inspection for Arkansas Nuclear One, Unit 1 (ANO-1), is the 1992 Edition. ASME Section XI, 1992 Edition, Table IWB-2500-1, Examination Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles, Item B5.10 requires a volumetric examination of the weld examination volume as identified in Figure IWB-2500-8.
On September 22, 1999, 10 CFR 50.55a was revised mandating the implementation of the 1995 Edition, through and including the 1996 Addenda of ASME Section XI, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems. These requirements are now implemented, as applicable, in the in-service inspection program for ANO, Unit 1.
The volumetric examination is to be conducted in accordance with ASME Section XI, Appendix VIII, Supplement 10.
II.
Relief Requested Entergy proposes to use Code Case N-695 with a root mean square error (RMSE) of 0.189 inches instead of the 0.125 inches specified for depth sizing in the Code Case. In the event an indication is detected that requires depth sizing, the 0.064-inch difference between the required RMSE and the demonstrated RMSE (0.189 inches - 0.125 inches
= 0.064 inches) will be added to the measured through-wall extent of the detected indication for comparison with the applicable acceptance criteria. If the examination Components/Numbers:
Core Flood Safe-End-to-Nozzle Welds01-025, 01-026 Code Classes:
ASME Code Class 1
References:
ASME Section XI, Division 1, 1992 Edition 1995 Edition with 1996 Addenda, Appendix VIII Examination Category:
B-F Item Number(s):
B5.10
==
Description:==
Reactor Vessel Nozzle-to-Safe End Butt Welds Unit / Inspection Interval Applicability:
Arkansas Nuclear One, Unit 1 / Third (3rd) 10-year interval, 1R21 Refueling Outage to 1CAN010801 Page 2 of 3 vendor demonstrates an improved depth sizing RMSE prior to the examination, the excess of that improved RMSE over the 0.125-inch RMSE requirement, if any, will be added to the measured value for comparison with applicable acceptance criteria.
The activities included in the relief request are subject to third party review by the Authorized Nuclear Inservice Inspector.
III. Basis for Alternative ASME Code Case N-695, Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1, is shown as acceptable for use in Regulatory Guide (RG) 1.147, Revision 14, dated August 2005.
To date, although examination vendors have qualified for detection and length sizing on these welds, the examination vendors have not met the RMSE requirement for depth sizing. Entergys contracted examination vendor has demonstrated ability to meet the depth sizing qualification requirement with an RMSE of 0.189 inches instead of the 0.125 inches required by the Code Case.
The addition of the difference in allowable depth sizing tolerance as demonstrated to that actually measured during the examination to the flaw depths measured will compensate for the possible variance in the measured depth.
The proposed alternative assures that the safe end-to-nozzle welds will be fully examined by procedures, personnel and equipment qualified by demonstration in all aspects except depth sizing. For depth sizing, the proposed addition of the difference between the qualified and demonstrated sizing tolerance to any flaw required to be sized compensates for the potential variation. The proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3).
IV. Conclusion 10 CFR 50.55a(a)(3) states:
Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g) and (h) of this section or portions thereof may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that:
(i)
The proposed alternatives would provide an acceptable level of quality and safety, or (ii)
Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The proposed alternative assures that the safe end-to-nozzle welds will be fully examined by procedures, personnel and equipment qualified by demonstration in all aspects except depth sizing. For depth sizing, the proposed addition of the difference between the qualified and demonstrated sizing tolerance to any flaw required to be sized to 1CAN010801 Page 3 of 3 compensates for the potential variation. Therefore, Entergy requests authorization to perform the proposed alternative to the Code requirement pursuant to 10 CFR 50.55a(a)(3) for implementation during the ANO-1, 1R21 refueling outage scheduled for the fall of 2008.
1CAN010801 Request for Relief ANO1-ISI-011 to 1CAN010801 Page 1 of 3 REQUEST FOR RELIEF ANO1-ISI-011 I.
Code Requirement(s)
The current code of record governing in-service inspection for Arkansas Nuclear One, Unit 1 (ANO-1), is the 1992 Edition. ASME Section XI, 1992 Edition, Table IWB-2500-1, Examination Category B-J, Pressure Retaining Welds in Piping, Item B9.11 requires a volumetric examination of essentially 100% of the weld length of the required examination volume as identified in Figure IWB-2500-8.
On September 22, 1999, 10 CFR 50.55a was revised mandating the implementation of the 1995 Edition, through and including the 1996 Addenda of ASME Section XI, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems. These requirements are now implemented, as applicable, in the in-service inspection program for ANO-1. The volumetric examination is to be conducted in accordance with ASME Section XI, Appendix XIII, Supplement 3.
Supplement 3 - Qualification Requirements for Ferritic Piping Welds - requires qualification of examination procedures, equipment, and personnel for ferritic pipe examination be accomplished by satisfying the requirements of Supplement 2, with the exception of the sample material and sample set defects.
Supplement 2 - Qualification Requirements for Wrought Austenitic Piping Welds -
provides the qualification requirements for ultrasonic examination that apply to austenitic welds, which are also applied to ferritic welds with the exceptions noted in Supplement 3.
Components/Numbers:
Reactor Vessel Outlet Nozzle to Pipe Welds and Pipe to Reactor Vessel Inlet Nozzle Pipe Welds14-028, 15-026,07-017, 09-015,11-015, 13-015 Code Classes:
ASME Code Class 1
References:
ASME Section XI, Division 1, 1992 Edition, Table IWB-2500-1 ASME Section XI, Division 1, 1995 Edition with 1996 Addenda, Appendix VIII Examination Category:
B-J Item Number(s):
B9.11
==
Description:==
Pressure Retaining Welds in Piping Unit / Inspection Interval Applicability:
Arkansas Nuclear One, Unit 1 / Third (3rd) 10-year interval, 1R21 Refueling Outage to 1CAN010801 Page 2 of 3 Paragraph 2.0, Conduct of Performance Demonstrations, states the specimen inside surface and identification shall be concealed from the candidate.
II.
Proposed Alternative Examinations Entergy proposes to use qualified vendor examination procedures, personnel and equipment that meet the qualification requirements of the 1995 Edition with 1996 Addenda of ASME Section XI, Appendix VIII, Supplements 2 and 3, with the exception that the inside surface of the pipe is not concealed as specified in paragraph 2 of Supplement 2.
The examination vendor procedures are qualified to meet the 2002 Addenda of ASME Section XI, Appendix VIII, Supplements 2 and 3, which specify requirements for qualification of scanning from the inside surface of piping.
The activities included in the relief request are subject to third party review by the Authorized Nuclear Inservice Inspector.
III. Basis for Alternative Ultrasonic examinations of piping welds connected to the reactor vessel nozzles are performed by qualified vendor personnel and procedures with an automated process from the inside of the vessel and associated piping. The qualification process described by the above referenced code requirements would prohibit access to the inside surface of the piping by the candidate by virtue of the inside surface being concealed from the candidate.
The vendor procedure and personnel qualification process for examination of piping welds from the inside surface is performed with the flaw location and specimen identification being obscured as to maintain a blind test.
The 2002 Addenda of ASME Section XI, Appendix VIII, Supplement 2 provide additional guidance for the conduct of performance demonstration activities when performed from the inside surface. However, 10 CFR 50.55a(b)(2)(xv) limits modification of Appendix VIII qualification requirements to the 1995 Edition through the 2001 Edition of ASME Section XI.
The proposed alternative assures that the ferritic piping welds will be fully examined by procedures, personnel and equipment qualified by demonstration in all aspects of the 1995 Edition, through and including the 1996 Addenda of ASME Section XI, Appendix VIII, as modified to allow the qualification of examinations from the inside surface as specified in the 2002 Addenda of ASME Section XI, Appendix VIII, Supplement 2.
to 1CAN010801 Page 3 of 3 IV. Conclusion 10 CFR 50.55a(a)(3) states:
Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g) and (h) of this section or portions thereof may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that:
(i)
The proposed alternatives would provide an acceptable level of quality and safety, or (ii)
Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
As discussed in Section III above, performing the examinations using vendor personnel and procedures that are qualified from the inside surface of the piping, taking direction from the 2002 addenda of ASME Section XI, Appendix VIII, Supplement 2, provides assurance that the piping welds are free of service related flaws and ensures plant safety and reliability. Therefore, Entergy requests authorization to perform the proposed alternative to the Code requirement pursuant to 10 CFR 50.55a(a)(3) for implementation during the ANO-1, 1R21 refueling outage scheduled for the fall of 2008.
1CAN010801 List of Regulatory Commitments to 1CAN010801 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
TYPE (Check one)
COMMITMENT ONE-TIME ACTION CONTINUING COMPLIANCE SCHEDULED COMPLETION DATE (If Required)
Entergy will use Code Case N-695 with a root mean square error (RMSE) of 0.189 inches instead of the 0.125 inches specified for depth sizing in the code case. In the event an indication is detected that requires depth sizing, the 0.064 inch difference between the required RMSE and the demonstrated RSME will be added to the measured through-wall extent of the detected indication.
X