ML093521350
| ML093521350 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/12/2010 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Minahan S Nebraska Public Power District (NPPD) |
| Lyon C Fred, NRR/DORL/LPL4, 301-415-2296 | |
| References | |
| TAC ME0687, TAC ME0688 | |
| Download: ML093521350 (11) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 12, 2010 Mr. Stewart B. Minahan Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321
SUBJECT:
COOPER NUCLEAR STATION - REQUESTS FOR RELIEF NO. RI-21 and RI-22 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL REGARDING VOLUMETRIC EXAMINATION COVERAGE FOR CERTAIN WELDS (TAC NOS. ME0687 AND ME0688)
Dear Mr. Minahan:
By letter dated February 16, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090540420), to the U.S. Nuclear Regulatory Commission (NRC), as supplemented by letter dated September 18, 2009 (ADAMS Accession No. ML092650190),
Nebraska Public Power District (the licensee) submitted requests for relief No. RI-21 and RI-22, from certain inservice inspection (lSI) requirements of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) at Cooper Nuclear Station (CNS). Specifically, the licensee requested in RI-21 to use an alternative examination volume defined in Code Case N-613-1, Figure 1, instead of the 100 percent volumetric examination of the reactor nozzle-to-vessel welds defined by the ASME Code. The licensee requested in RI-22 to perform examinations of accessible portions of the reactor nozzle-to-safe end welds to the extent practical instead of the examinations required by the ASME Code. The applicable ASME Code at CNS for the fourth 10-year lSI interval, which commenced on March 1, 2006, is the 2001 Edition through the 2003 Addenda.
The requests for relief were proposed pursuant to the provisions of paragraph 50.55a(g)(6)(i) of Title 10 of the Code of Federal Regulations (10 CFR). Based on the information you provided in your requests for relief, the NRC staff determined that the ASME Code requirement is impractical and that reasonable assurance of structural integrity of the subject components has been provided by the examinations performed. Granting the requests for relief is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, requests for relief RI-21 and RI-22 are granted pursuant to 10 CFR 50.55a(g)(6)(i) for the fourth 10-year lSI 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.
S. Minahan
- 2 The detailed results of the NRC staff review are provided in the enclosed safety evaluation. If you have any questions concerning this matter, please contact Mr. F. Lyon of my staff at (301) 415-2296 or via e-mail at fred.lyon@nrc.gov.
Sincerely, Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL REQUESTS FOR RELIEF NO. RI-21 AND RI-22 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298
1.0 INTRODUCTION
By letter dated February 16, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090540420), to the U.S. Nuclear Regulatory Commission (NRC), as supplemented by letter dated September 18, 2009 (ADAMS Accession No. ML092650190),
Nebraska Public Power District (the licensee) submitted requests for relief No. RI-21 and RI-22, from certain inservice inspection (lSI) requirements of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) at Cooper Nuclear Station (CNS). Specifically, the licensee requested in RI-21 to use an alternative examination volume defined in Code Case N-613-1, Figure 1, instead of the 100 percent volumetric examination of the reactor nozzle-to-vessel welds defined by the ASME Code. The licensee requested in RI-22 to perform examinations of accessible portions of the reactor nozzle-to-safe end welds to the extent practical instead of the examinations required by the ASME Code.
2.0 REGULATORY EVALUATION
The lSI of the ASME Code Class 1, 2, and 3 components is to be performed in accordance with the applicable edition and addenda of ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," as required by Title 10 of the Code of Federal Regulations, Part 50 (specifically 10 CFR 50.55a(g)), except where specific relief has been granted by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). Paragraph 10 CFR 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the applicant demonstrates that: (i) the proposed alternatives 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.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the Enclosure
- 2 components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1O-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The Code of record for the fourth 10-year lSI interval at CNS is the ASME Code,Section XI, 200'1 Edition with 2003 Addenda.
3.0 RELIEF REQUEST RI-21 3.1 Components for Which Relief is Requested American Society of Mechanical Engineers (ASME), Class 1, Reactor Pressure Vessel (RPV)
Full Penetration Nozzle-to-Vessel Welds Examination Category B-O, Full Penetration Welded Nozzles in Vessels, Item Number B3.90 in Table IWB-2500-1 of Section XI of the ASME Code, requires 100 percent volumetric examination of each weld.
Component no. and description Percentage of weld examined by ultrasonic testing, 3rd Interval Percentage of weld examined by ultrasonic testing, 4th Interval NVE-BO-N1A Recirculation Inlet 32 46 NVE-BO-N2E Recirculation Outlet 40 55 NVE-BO-N2H Recirculation Outlet 40 55 NVE-BO-N2K Recirculation Outlet 40 55 NVE-BO-N3A Main Steam 35 49 NVE-BO-N5A Core Spray 31 51 3.2 Applicable ASME Code Edition and Addenda The Code of record for the fourth 1O-year lSI interval is the ASME Code,Section XI, 2001 Edition, with 2003 Addenda.
3.3 Licensee's Proposed Alternative Examination (as stated by licensee)
In lieu of performing the [ASME Code]-required examinations, CNS proposes to examine the accessible portions of reactor vessel nozzle-to-vessel welds to the extent practical.
- 3 3.4 Impracticality of Compliance (as stated by licensee)
Pursuant to 10 CFR 50.55a(g)(5)(iii), Nebraska Public Power District has determined that compliance with the [ASME Code] requirements of achieving essentially 100% coverage of the welds listed in Table RI-21-1 [1J is impractical for CNS. The CNS construction permit was issued before the effective date of implementation for ASME [Code] Section XI, thus the plant was not designed to meet the requirements of inservice inspection. The configuration of the nozzles, the design of the vessel insulation support rings and the nozzle access hatches, and interferences from thermocouple pads, instrument lines, etc., prevent 100%
examination of the required weld volumes.
[Also, t]he nozzle geometry itself limits the physical access of the ultrasonic examination (UT) probes to only single sided access.... The UT examinations can only be performed from the vessel side of the nozzle to vessel weld as no current qualified techniques exist for performing UT examinations from the nozzle bore or reactor vessel inside diameter (lD) in order to achieve >90% coverage.
3.5 Burden Caused by Compliance (as stated by licensee)
A major modification to the reactor vessel nozzles and access hatches would be required....
[This would include a c]omplete redesign and replacement of the reactor vessel nozzles [ ] to provide enough clearance so the nozzle to vessel weld could be examined from the nozzle side as well as the vessel side. Such a replacement is considered impractical due to the significant dose and extensive outage time needed to complete these changes without a compensating increase in safety for an installed reactor.
Therefore, obtaining essentially 100 percent coverage is not feasible or practical.
3.6 Licensee's Basis for Requesting Relief (as stated by licensee)
The examinations were performed in accordance with Appendix VIII of ASME
[Code] Section XI using qualified personnel, procedures and equipment with the applicable limitations set forth in 10 CFR 50.55a. The extent of the nozzle restrictions and the total volume accessible for examination, based on the examinations performed in the fourth interval are compared to examinations performed in the previous third interval. In all cases, the coverage achieved in the fourth interval exceeded the coverage achieved in the previous interval.
The proposed alternative recognizes the limitations in the nozzle geometry that affect examination coverage. The inner 15% of the nozzle to vessel weld (Le.,
the vessellD side) was fully interrogated which is the key area of interest for 1 Table RI-21-1 refers to a table in the licensee's February 16, 2009, submittal which is not reproduced here.
-4 service induced flaws initiated on the inside diameter of the weld and heat affected zone. No indications were identified providing reasonable assurance that continued structural integrity will be maintained.
The nozzle-to-shell welds in this request are visually examined (VT-2) during the ASME
[Code] Section XI Class 1 system pressure test each refueling outage. The VT-2 examinations performed as part of the reactor vessel pressure test also did not detect any leaks in these welds.
3.7
NRC Staff Evaluation
The ASME Code requires 100 percent volumetric examination of the weld length of the welds shown in 3.1 above. The licensee noted that the interferences from these items preclude the complete UT examination of these welds. The licensee has examined a portion of the subject welds (see percent of weld examined above). Also, the licensee found no recordable indications by its best effort volumetric examination or by the visual examinations performed. The NRC staff determined through the review of the data provided by the licensee that the ASME Code requirements are impractical and design modifications would be necessary to provide access for examination to meet the ASME Code requirements. Imposition of the ASME Code requirements would result in an undue burden on the licensee. Based on the licensee's examinations, reasonable assurance of structural integrity of the subject components has been provided; therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
4.0 RELIEF REQUEST RI-22 4.1 Components for Which Relief is Requested American Society of Mechanical Engineers (ASME), Class 1, RPV Nozzle-to-Safe End Butt Welds Examination Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles Item Number B5.1 0, in Table IWB-2500-1, of Section XI of the ASME Code, requires 100 percent volumetric and surface examination of the welds.
Component no. and Description Percentage of weld examined by ultrasonic testing, 4th Interval CS-BF-Ix Core Spray 75 RRE-BF-1 Recirculation Outlet 75 RRH-BF-1 Recirculation Outlet 75 RRK-BF-1 Recirculation Outlet 75 4.2 Applicable ASME Code Edition and Addenda The Code of record for the fourth 10-year lSI interval is ASME Code Section XI, 2001 Edition, with 2003 Addenda.
- 5 4.3 Licensee's Proposed Alternative Examination (as stated by licensee)
In lieu of performing the [ASME Code]-required examinations, CNS proposes to examine the accessible portions of reactor vessel nozzle-to-safe-end welds to the extent practical.
4.4 Impracticality of Compliance (as stated by licensee)
Pursuant to 10 CFR 50.55a(g)(5)(iii), Nebraska Public Power District has determined that compliance with the [ASME Code] requirements of achieving essentially 100% coverage of welds listed in Table RI-22-1 [2] is impractical....
CNS replaced the Class 1 stainless steel piping in the mid-1980's. The best techniques available to mitigate intergranular stress corrosion cracking (IGSCC) were used: IGSCC resistant material, corrosion resistant cladding on the internal diameter of the nozzle buttering, and Induction Heat Stress Improvement (IHSI).
These activities provided welds that are Generic Letter 88-01 IGSCC Category "A." The resultant welds were examined using the standard industry techniques for IGSCC at that time. Typical vOlumetric examination coverage was previously reported to be greater than 90%. The radial shrinkage caused by the IHSI in the heat affected zone areas during the construction of the welds prevents full contact of the ultrasonic testing (UT) probes.
4.5 Burden Caused by Compliance (as stated by licensee)
Removal of the nozzle base metal by grinding or machining would be required to obtain a surface conducive for examination. Therefore, obtaining essentially 100% percent coverage is not feasible or practical.
Examination coverage was limited due to weld shrinkage in the heat affected zone on the nozzle resulting in a greater than 1/32" gap between the search unit and examination surface. As stated [above], no external weld conditioning is possible without removing the nozzle base material in order to achieve a smooth surface needed for the additional circumferential scan that could not be performed thus limiting the examination coverage to 75%. As an alternative, a modification that would improve examination coverage by improving the nozzle weld surface would be to install a weld overlay subject to NRC approval. The installation of this repair method to improve weld coverage is impractical as the significant cost and estimated dose to install this type of modification is not commensurate with the incremental increase in safety.
4.6 Licensee's Basis for Requesting Relief (as stated by licensee)
The examinations were performed in accordance with Appendix VIII of ASME
[Code] Section XI using qualified personnel, procedures and equipment with the 2 Table RI-22-1 refers to a table in the licensee's February 16, 2009, submittal which is not reproduced here.
- 6 applicable limitations set forth in 10 CFR 50.55a. The current Performance Demonstration Initiative (POI) techniques that implement Appendix VIII of ASME
[Code] Section XI are more restrictive on the requirements for weld profiles. The current examinations are considered more reliable than previously performed volumetric examinations.
The weld profile shown in Figure RI-22-1 [3], indicates that the shrinkage in the heat affected zone on the nozzle side caused a greater than 1/32" gap between the UT search unit and examination surface. Both axial UT scans obtained 100%
coverage necessary to detect circumferential flaws. However, the circumferential scans necessary to detect axial flaws could not be fully obtained. The circumferential scan on the safe-end was obtained but the circumferential scan on the nozzle side could not be obtained thus resulting in the combined volumetric coverage estimates provided in [4.1 above].
The examination coverage achieved for the required examination volume of ASME [Code] Section XI Figure IWB-2500-8 and the CNS Risk-Informed Program was obtained as follows:
50% [ASME] Code volume based on achieving 100% coverage of the two required axial scans for the detection of circumferentially oriented flaws 25% [ASME] Code volume based on achieving 100% coverage of one the two required circumferential scans for the detection of axially oriented flaws The coverage calculation is based on obtaining full coverage on three of the four scans:
Axial scan coverage (50%) + Circumferential scan coverage (50%/2 =
25%) = 75%
The [UT] examination[s] [were] able to interrogate the root area and the heat affected zones in the inner 1/3 of the weld[s]. This is the primary area of concern for service induced intergranular stress corrosion cracking (IGSCC) flaws. These welds along with the safe-ends were replaced in the 1984-85 outage with material resistant to IGSCC using Inconel-82 welds that included an Inconel-82 corrosion resistant cladding over the existing Inconel-182 weld butter, and a secondary mitigation method of Induction Heating Stress Improvement was applied. 100%
coverage of the two axial scans did not detect any circumferentially-oriented flaws. 100 percent coverage of one of the two required circumferential scans for the detection of axially-oriented flaws did not detect any flaws. The examinations performed plus the IGSCC resistance of these welds provides reasonable assurance of the continued structural integrity of these welds.
3 Figure RI-22-1 refers to a figure from the licensee's February 16, 2009, submittal which is not reproduced here.
- 7 The nozzle-to-safe end welds in this request are visually examined during the system pressure test each refueling outage Surface examinations are not required in accordance with the CNS Risk-Informed lSI Program. The VT-2 examinations performed as part of the reactor vessel pressure test also did not detect any leaks in these welds.
4.7
NRC Staff Evaluation
The ASME Code requires 100 percent volumetric examination of the welds shown in 4.1 above.
The licensee noted that the weld shrinkage in the heat-affected zone on the nozzle side of the weld caused a greater than 1/32" gap between the UT search unit and examination surface.
Both axial UT scans obtained 100 percent coverage necessary to detect circumferential flaws.
However, the circumferential scans necessary to detect axial flaws could not be fully obtained.
The circumferential scan on the safe-end was obtained but the circumferential scan on the nozzle side could not be obtained thus resulting in the 75 percent volumetric coverage. The licensee has examined a portion of the subject welds and found no recordable indications by its best effort volumetric examination or by the visual examinations performed.
Removal of the nozzle base metal by grinding or machining would be required to obtain a surface conducive for examination. Therefore, obtaining essentially 100 percent coverage is not feasible or practical. Another modification that would improve examination coverage by improving the nozzle weld surface would be to install a weld overlay subject to NRC approval.
The installation of these repair methods to improve weld coverage is impractical as the significant cost and estimated dose to install this type of modification is not commensurate with the incremental increase in safety.
The NRC staff determined through the review of the data provided by the licensee that the ASME Code requirements are impractical and weld modifications would be necessary to meet the ASME Code requirements. Imposition of the ASME Code requirements would result in an undue burden on the licensee. Based on the licensee's examinations, reasonable assurance of structural integrity of the subject components has been provided; therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
5.0 CONCLUSION
The NRC staff has reviewed the licensee's submittal and concludes that ASME Code examination coverage requirements are impractical for the subject welds listed in Relief Requests RI-21 and RI-22. Furthermore, imposition of these ASME Code requirements would create a burden on the licensee. The staff further determined that based on the VOlumetric and visual coverage obtained on the subject welds, it is reasonable to conclude that if significant service-induced degradation had occurred, evidence of it would have been detected by the examinations that were performed. Furthermore, the staff concluded that the examinations, performed to the extent practical, provide reasonable assurance of structural integrity of the subject welds.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the NRC staff grants
- 8 relief for the subject examinations of the components contained in RI-21 and RI-22 at CNS for the fourth 10-year lSI interval.
The staff has further determined that granting Relief Requests RI-21 and RI-22 pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor:
E. Andruszkiewicz Date: January 12, 2010
- memo dated OFFICE NRR/LPL4/PM NRR/LPL4/LA DCI/CVIB/BC NRR/LPL4/BC NRR/LPL4/PM NAME FLyon JBurkhardt MMitchell*
MMarkley FLyon DATE 1/8/10 1/8/10 12/14/09 1/11/10 1/12/10