ML042160462
| ML042160462 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 08/03/2004 |
| From: | Gramm R NRC/NRR/DLPM/LPD4 |
| To: | Edington R Nebraska Public Power District (NPPD) |
| Honcharik J, NRR/DE/EMCB, 301-415-1157 | |
| References | |
| TAC MC2058, TAC MC2059, TAC MC2060 | |
| Download: ML042160462 (11) | |
Text
August 3, 2004 Mr. Randy K. Edington Vice President-Nuclear and CNO Nebraska Public Power District P.O. Box 98 Brownville, NE 68321
SUBJECT:
COOPER NUCLEAR STATION - RE: REQUESTS PR-03, RC-06, AND RC-07 FOR RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) BOILER AND PRESSURE VESSEL CODE (CODE) (TAC NOS. MC2058, MC2059, AND MC2060)
Dear Mr. Edington:
By letter dated February 12, 2004, as supplemented by two letters dated May 18, 2004, Nebraska Public Power District (the licensee) submitted requests for the use of alternatives to certain ASME Code Section XI requirements at Cooper Nuclear Station (CNS).
The proposed alternatives for relief requests RC-06 and RC-07 provide an acceptable level of quality and safety. Therefore, pursuant to Section 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (10 CFR), the NRC staff authorizes the use of relief requests RC-06 and RC-07 during the licensees first containment inservice inspection. Pursuant to the regulation at 10 CFR 50.55a(g)(4)(iv), the NRC staff approves the use of a later code edition and addenda for inservice inspection items as requested in PR-03. The NRC staff's safety evaluation for RC-06, RC-07, and PR-03 is enclosed.
Sincerely,
/RA/
Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosure:
Safety Evaluation cc w/encl: See next page
August 3, 2004 Mr. Randy K. Edington Vice President-Nuclear and CNO Nebraska Public Power District P.O. Box 98 Brownville, NE 68321
SUBJECT:
COOPER NUCLEAR STATION - RE: REQUESTS PR-03, RC-06, AND RC-07 FOR RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) BOILER AND PRESSURE VESSEL CODE (CODE) (TAC NOS. MC2058, MC2059, AND MC2060)
Dear Mr. Edington:
By letter dated February 12, 2004, as supplemented by two letters dated May 18, 2004, Nebraska Public Power District (the licensee) submitted requests for the use of alternatives to certain ASME Code Section XI requirements at Cooper Nuclear Station (CNS).
The proposed alternatives for relief requests RC-06 and RC-07 provide an acceptable level of quality and safety. Therefore, pursuant to Section 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (10 CFR), the NRC staff authorizes the use of relief requests RC-06 and RC-07 during the licensees first containment inservice inspection. Pursuant to the regulation at 10 CFR 50.55a(g)(4)(iv), the NRC staff approves the use of a later code edition and addenda for inservice inspection items as requested in PR-03. The NRC staff's safety evaluation for RC-06, RC-07, and PR-03 is enclosed.
Sincerely,
/RA/
Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
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7/27/04 7/29/04 OFFICIAL RECORD COPY
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NOS. RC-06, RC-07, AND PR-03 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298
1.0 INTRODUCTION
By letter dated February 12, 2004 (Reference 1) and two supplements dated May 14, 2004 (References 2 and 3), Nebraska Public Power District (the licensee, NPPD) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI (Reference 4) for Cooper Nuclear Station (CNS).
Relief request PR-03 is applicable to the third 10-year inservice inspection (ISI) interval, which ends February 28, 2006. Relief requests RC-06 and RC-07, related to (1) the alternative requirements for IWE-5240 visual examination and (2) the requirements of IWE-3515.1 for examination of pressure retaining bolts, are applicable to the first 10-year interval of the Containment Inspection Program which ends April 8, 2007.
2.0 REGULATORY EVALUATION
The regulation at Section 50.55a(g)(4) of Title 10 of the Code of Federal Regulations (10 CFR) requires licensees to update their ISI program plan every 10 years using the latest edition and addenda of the ASME Code for ISI of Class 1, Class 2, Class 3, Class MC, and Class CC components. Accordingly, in preparing for its third ISI program plan, the licensee recognized the need for relief from some of the requirements in Reference 4 as incorporated by reference in 10 CFR 50.55a. The reliefs are requested pursuant to 10 CFR 50.55a(a)(3) and 10 CFR 50.55a(g)(4).
3.0 EVALUATION OF RELIEF REQUEST RC-06 Perform a VT-1 or VT-3 visual examination in lieu of a VT-2 visual examination following a repair or replacement.
The Items for which Relief is Requested:
Code Class:
MC
Reference:
IWE-5240 Examination Category:
E-P Item
Description:
VT-2 Visual Examination Component Numbers:
All Code Requirement:
IWE-5240 states that the requirements of IWA-5240 are applicable following repair, replacement, or modification. IWA-5240 requires a VT-2 visual examination in conjunction with the pressure test.
Licensees Proposed Alternative (as stated in Reference 1):
In lieu of performing a VT-2 examination for repair or replacement, a VT-1 or VT-3 examination, as appropriate, will be performed.
Licensees Basis for Relief (as stated Reference 1):
Subsubarticle IWE-5210 states that except as noted in Subsubarticle IWE -5240, the requirements of Article IWA-5000 are not applicable to Class MC or Class CC components. Subsubarticle IWE-5240 states that the requirements of Paragraph IWA-5246 (corrected to IWA-5240 in the 1993 Addenda) for visual examinations are applicable. Subsubarticle IWA-5240 identifies requirements for the performance of a VT-2 visual examination. VT-2 visual examinations are conducted to detect evidence of leakage from pressure-retaining components with or without leakage collection systems, during the conduct of a system pressure test. In addition, personnel performing VT-2 visual examination are required to be qualified in accordance with Subarticle IWA-2300 of ASME Section XI.
Table IWE-2500-1, Examination Category E-P, Item E9.10, identifies the examination method of 10CFR50 Appendix J and does not identify a VT-2 visual examination.
10CFR50, Appendix J provides requirements for testing, as well as acceptable leakage criteria. These tests are performed by qualified Appendix J test personnel using calibrated equipment to determine leak rate acceptability.
The 1998 Edition of Section XI, Subsubarticle IWE-5240, requires a detailed visual examination (IWE-2310) be performed on areas affected by repair/replacement activities. The requirement was amended by 10CFR50.55a (b)(2)(ix)(F) to require a VT-1 or VT-3 examination.
Repairs and replacements, including modification, to the containment pressure-retaining boundary and to integral attachments, must be performed in accordance with Article IWA-4000. This article requires, among other things, preparation of a repair and replacement plan; requires repairs and installation of replacement, including performance of nondestructive examinations, to be performed in accordance with the original edition of the Construction Code or Section XI; and requires performance of preservice inspection in accordance with Subsection IWE. The program specifies the repair methods and nondestructive examinations necessary to ensure that the original quality and construction requirements of the containment vessel are met.
Performance of the Appendix J testing will detect leakage that may exist in the containment pressure-retaining boundary. In accordance with the requirements of the 1998 Edition of Section XI, Subsubarticle IWE-5240, performance of a VT-1 or VT-3 examination (instead o[f] a VT-2 examination) and compliance with Article IWA-4000 will provide assurance of the structural integrity of the containment pressure-retaining boundary.
Relief is requested in accordance with 10CFR50.55a(a)(3)(i).
Evaluation:
The NRC staff has previously granted relief from the requirements of Subsubarticle IWE-5240 of Subsection IWE of Reference 4. The basis for relief had been either: (1) if a licensee commits to perform VT-1 examination of the repaired areas during the appropriate pressure test, or (2) if a licensee commits to the provisions of ASME Code Case N-649. For CNS, the licensee requested relief from the IWE-5241 requirement. However, the licensee did not propose any of the above alternatives. Instead, the licensee has proposed: In lieu of performing a VT-2 examination for repair or replacement, a VT-1 or VT-3 examination, as appropriate, will be performed. As the proposed alternative does not address certain specificity as to when the licensee would use VT-1 and VT-3 examination, the NRC staff requested additional information as follows:
(a)
The intent of IWE-5240 is to detect the areas of leakage during pressure testing of the containment pressure boundary components. How would you perform the examination of a leaking component which is in part inaccessible to detect leakage area(s)
(e.g., Type B testing of vent pipes between the drywell and the suppression chamber)?
In Reference 2, the licensee stated that: If an area that is normally inaccessible requires a repair or replacement, the area would be made accessible to perform the repair or replacement (e.g., removal of the guard pipe or the concrete). The required examination, whether it is a VT-2, VT-3 or VT-1, would be performed prior to the full restoration of the configuration (i.e., reinstallation of the guard pipe or concrete). The NRC staff finds the process stated in the licensees response acceptable.
(b)
In some cases, IWE-5222 permits deferral of pressure test following minor repair. How would you perform the examination of the repaired areas to ensure that the repaired areas will not be leaking during the deferral period?
In Reference 2, the licensee stated: NPPD is not requesting relief from the provisions of IWE-5222. IWE-5222 allows deferral of the leakage test for minor repairs or modifications to the pressure retaining boundary, provided nondestructive testing is performed in accordance with the approved repair program. The nondestructive testing in accordance with an approved repair program will provide reasonable assurance that the repaired area is not leaking until the next scheduled leak test is performed.
For certain minor repairs as indicated in IWE-5222, the Working Group on Containments, that developed the IWE requirements, considered the use of detailed visual examination as adequate nondestructive examination (NDE), as reflected in the 1998 Edition of the ASME Code. The response indicates that the licensee will perform appropriate NDE to confirm the integrity of the minor repair areas. Therefore, the NRC staff finds that the licensees approach provides an acceptable level of quality and safety.
The response to the request for additional information indicates that the licensee has developed a formal procedure for addressing various situations during and after repair and replacement activities, in lieu of only VT-2 examination as required by the IWE-5240 of the 1992 Edition and the 1992 Addenda of the ASME Code. Therefore, the NRC staff finds that the proposed alternative is acceptable, because it provides an acceptable level of quality and safety.
4.0 EVALUATION OF RELIEF REQUEST RC-07 In lieu of using the acceptance criteria provided in IWE-3515.1, use the acceptance criteria in Subarticle IWB-3517.1.
The Items for which Relief is Requested:
Code Class:
MC
Reference:
IWE-3515.1 Examination Category:
E-G Item
Description:
Bolted Connections Component Numbers:
All Code Requirement:
IWE-3515.1 requires that "bolting material shall be examined in accordance with the material specification for defects which may cause the bolted connection to violate either the leak-tight or structural integrity. Defective items shall be replaced."
Licensees Proposed Alternative (as stated in Reference 1):
Bolting material will be examined in accordance with the inservice standards of the 1992 Edition, with 1992 Addenda of ASME Section XI, Subparagraph IWB-3517.1 Licensees Basis for Relief (as stated in Reference 1):
Bolting material specifications provide requirements relative to the base material properties and related fabrication discontinuities. Material specification requirements are generally associated with the chemical composition, mechanical strength, test specimens and dimensional requirements. These qualities provide little or no guidance for the examination of the bolted connection for service-induced degradation. For inservice bolting, examination guidelines and acceptance criteria must be specific to discontinuities which are relevant to continued service.
In lieu of using the acceptance criteria provided in IWE-3515.1, NPPD proposes to use the acceptance criteria for Class 1 pressure retaining bolting. Subparagraph IWB-3517.1, Standards for Examination Category B-G-1, Pressure Retaining Bolting Greater Than 2 in. in Diameter, and Examination Category B-G-2, Pressure Retaining Bolting 2 in. and Less in Diameter.
The following relevant conditions shall require correction to meet the requirements of IWB-3122 prior to service or IWB-3142 prior to continued service; a) crack-like flaws that exceed the allowable liner flaw standards of IWB-3515.5; b) more than one deformed or sheared thread in zone of thread engagement of bolts, studs or nuts; c) localized general corrosion that reduces the bolt or stud cross-sectional area by more than 5%;
d) bending, twisting, or deformation of bolts or studs to the extent that assembly or disassembly is impaired; e) missing or loose bolts, studs, nuts, or washers; f) fractured bolts, studs, or nuts; g) degradation of protective coatings on bolting surfaces; or h) evidence of coolant leakage near bolting.
Relief is requested in accordance with 10CFR50.55a(a)(3)(i).
Evaluation:
IWE-3515.1 acceptance criteria for containment pressure-retaining bolting is non prescriptive, and depends upon the subjective criteria such as the extent of degradation that may violate the leak tightness or structural integrity of the bolted connection. In the revision to 10 CFR 50.55a(b)(xi)(H) dated September 26, 2002 (67 FR 60520), while incorporating by reference the 1997 Addenda through the 2000 Addenda of Section XI of the ASME Code, the Commission invoked the use of IWB-3517.1 for evaluating flaws and degradation in the containment pressure-retaining bolting. The licensee is pro-active in proposing this alternative in lieu of the vague requirement of IWE-3515.1 of the 1992 Edition of Subsection IWE. Therefore, the NRC staff finds that the proposed alternative is acceptable, because it provides an acceptable level of quality and safety.
5.0 EVALUATION OF RELIEF REQUEST PR-03 Eliminate the removal of bolting, VT-3 visual examination for corrosion, and IWA-3100 evaluation, on leaking bolted connections.
The Items for which Relief is Requested:
Code Classes:
1, 2, and 3
References:
IWA-5250 Examination Categories:
B-P, C-H, and D-A Item Numbers:
B15.10 through B15.71, C7.10 through C7.80, and D1.10
==
Description:==
Alternate corrective measures for bolted connections.
Component Numbers:
All Class 1, 2, and Class 3 pressure retaining components subject to system pressure testing Code Requirement:
IWA-5250(a)(2) requires, if leakage occurs at a bolted connection, that the bolting be removed, examined by VT-3 visual examination for corrosion, and evaluated in accordance with IWA-3100.
Licensees Proposed Alternative (as stated in Reference 3):
In lieu of the requirements contained in Subsubarticle IWA-5250 of the 1989 Edition of Section XI, NPPD will comply with the corrective action requirements for bolted connections contained in Subsubarticle IWA-5250 of the 1998 Edition, 1999 Addenda of Section XI.
Licensees Basis for Relief (as stated in Reference 3):
In the event of a bolted connection leak detected during the conduct of a system pressure test, the 1989 Edition, ASME Section XI requirements specify that all bolting must be removed for the purpose of a VT-3 visual examination and evaluation. This requires removing the component or piping system from service, which could result in a plant shutdown, a delay of plant startup or, for continued operation, a reduction in plant safety.
Nebraska Public Power District's (NPPD) proposed alternative to requirements of the 1989 Edition of the Code was approved by the Nuclear Regulatory Commission (NRC).
Specifically, NPPD would evaluate the bolting to determine its susceptibility to corrosion, perform a more in-depth evaluation as applicable and if necessary, remove the bolt closest to the source of leakage and evaluate the bolt in accordance with IWA-3100(a).
Since the granting of the above relief, the Section XI Code requirements have changed making clear the purpose of the examination is to detect degradation of bolting due to leakage from borated systems. The ASME Code Section XI, 1998 Edition, 1999 Addenda, revised Subsubarticle IWA-5250 (a)(2) as follows:
"If leakage occurs at a bolted connection in a system borated for the purpose of controlling reactivity, one of the bolts shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100. The bolt selected shall be one closest to the source of leakage. When the removed bolt has evidence of degradation, all remaining bolts in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100."
Under the revised subsubarticle, the examination is applicable to code piping of borated water systems. Cooper Nuclear Station (CNS) is a boiling water reactor and does not use borated water to control reactivity during normal plant operation.
No other changes made in the 1999 Addenda are related to the change made in Subsubarticle IWA-5250. System pressure testing requirements contained in Examination Tables IWB-2500-1, IWC-2500-1 and IWD-2500-1 were not affected.
Repair and replacement activities must still be performed in accordance with the repair and replacement program, as applicable.
NPPD requests approval to use the corrective action requirements of ASME Section XI, 1998 Edition, 1999 Addenda, Subsubarticle IWA-5250 in accordance with 10CFR50.55a(g)(4)(iv).
Evaluation:
The regulation at 10 CFR 50.55a(g)(4)(iv), states that the inservice examination of components and system pressure tests may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modifications listed therein and NRC approval. Portions of the editions or addenda may be used provided that all related requirements of the respective editions and addenda are met.
The licensees request to use the corrective action requirements for bolted connections contained in Subarticle IWA-5250 of the 1998 Edition, 1999 Addenda Section XI of the Code is approved pursuant to 10 CFR 50.55a(g)(4)(iv), for the third 10-year interval of the ISI program for CNS.
6.0 CONCLUSION
S On the basis of the information provided in the relief requests, and in the response to the request for additional information, the NRC staff concludes that the proposed alternatives to the requirements of IWE-5240 and IWE-3515.1 will provide acceptable level of quality and safety.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the use of relief requests RC-06 and RC-07 during the licensees first containment ISI at CNS.
Additionally, pursuant to the regulation at 10 CFR 50.55a(g)(4)(iv), the NRC staff approves the use of a later code edition and addenda for ISI items as requested in PR-03.
All other requirements of the ASME Code,Section XI, for which relief has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
7.0 REFERENCES
1.
Letter from R. Edington (NPPD) to NRC, Inservice Inspection Relief Requests PR-03, RC-06, RC-07, RI-17, RI-31, RI-32 and RI-33, February 12, 2004. ADAMS accession no. ML040490589.
2.
Letter from R. Edington (NPPD) to NRC, Response to Requests for Additional Information Regarding Relief Request RC-06, May 18, 2004. ADAMS accession no. ML041450302.
3.
Letter from R. Edington (NPPD) to NRC, Resubmittal of Inservice Inspection Relief Request PR-03, May 18, 2004. ADAMS accession no. ML041460314.
4.
Section XI of the ASME Boiler & Pressure Vessel Code, Rules for Inservice Inspection of Nuclear Power Plant Components, 1992 Edition and 1992 Addenda.
Principal Contributor: H. Ashar M. Honcharik Date: August 3, 2004
June 2004 Cooper Nuclear Station cc:
Mr. William J. Fehrman President and Chief Executive Officer Nebraska Public Power District 1414 15th Street Columbus, NE 68601 Mr. Clay C. Warren Vice President of Strategic Programs Nebraska Public Power District 1414 15th Street Columbus, NE 68601 Mr. John R. McPhail, General Counsel Nebraska Public Power District P. O. Box 499 Columbus, NE 68602-0499 Mr. Paul V. Fleming Licensing Manager Nebraska Public Power District P.O. Box 98 Brownville, NE 68321 Mr. Michael J. Linder, Director Nebraska Department of Environmental Quality P. O. Box 98922 Lincoln, NE 68509-8922 Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 1824 N Street Auburn, NE 68305 Ms. Cheryl K. Rogers, Program Manager Nebraska Health & Human Services System Division of Public Health Assurance Consumer Services Section 301 Centennial Mall, South P. O. Box 95007 Lincoln, NE 68509-5007 Mr. Ronald A. Kucera, Director of Intergovernmental Cooperation Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 218 Brownville, NE 68321 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 Jerry Uhlmann, Director State Emergency Management Agency P. O. Box 116 Jefferson City, MO 65101 Chief, Radiation and Asbestos Control Section Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson Suite 310 Topeka, KS 66612-1366 Mr. Daniel K. McGhee Bureau of Radiological Health Iowa Department of Public Health 401 SW 7th Street Suite D Des Moines, IA 50309 Mr. Scott Clardy, Director Section for Environmental Public Health P.O. Box 570 Jefferson City, MO 65102-0570 June 2004 Jerry C. Roberts, Director of Nuclear Safety Assurance Nebraska Public Power District P.O. Box 98 Brownville, NE 68321