ML110600356
| ML110600356 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/29/2011 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Conway J Pacific Gas & Electric Co |
| Polickoski J, NRR/DORL/LPL4, 415-5430 | |
| References | |
| TAC ME4577 | |
| Download: ML110600356 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 29, 2011 Mr. John T. Conway Senior Vice President - Energy Supply and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant 77 Beale Street, Mail Code B32 San Francisco, CA 94105 SUB,JECT: DIABLO CANYON POWER PLANT, UNIT NO.2 - APPROVAL OF REQUEST FOR RELIEF NDE-RCS-SE-2R16 FROM EXAMINATION REQUIREMENTS OF ASME CODE, SECTION XI, APPENDIX VIII, SUPPLEMENT 10, ROOT MEAN SQUARE ERROR (TAC NO. ME4577)
Dear Mr. Conway:
By letter dated August 12, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102350309), Pacific Gas and Electric Company (PG&E, the licensee) submitted a request for relief NDE-RCS-SE-2R16 from certain examination requirements of the American Society of Mechanical Engineers (AS ME) Boiler and Pressure Vessel Code (Code) at the Diablo Canyon Power Plant, Unit 2 (DCPP, Unit 2). Specifically, the licensee proposed using a root mean square error (RMSE) criterion for sizing flaws that is greater than ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds" (N-695). The licensee also requested relief from the N-695 exclusion for examinations through corrosion resistant clad. ASME Code Case N-695 is referenced in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," October 2007 (ADAMS Accession No. ML072070419)
The NRC staff completed its review of request for relief NDE-RCS-SE-2R16 and enclosed is the NRC staff's safety evaluation (SE). The licensee requested this request for relief pursuant to paragraph 50.55a(g)(5)(iii) of Title 10 of the Code of Federal Regulations (10 CFR). The NRC staff concluded that compliance with the N-695-required 0.125-inch RMSE criteria, at this time, is impractical, and the proposed alternative to add the difference between 0.189-inch RMSE and the ASME Code-required value (0.189-inch minus 0.125-inch = 0,064-inch) provides reasonable assurance of the structural integrity of the dissimilar metal welds that will be examined during the 16th refueling outage (2R16) which is in the third 10-year inservice inspection (lSI) interval.
The 16th refueling outage is scheduled to begin May 2011. The NRC staff also concluded that the vendor-performed, cladded versus uncladded equivalency demonstrations provided reasonable expectation that flaws from cladded surfaces would be detected using the vendor's demonstrated procedures and personnel.
Therefore, request for relief NDE-RCS-SE-2R16 is granted pursuant to 10 CFR 50.55a(g)(6)(i),
for the DCPP, Unit 2, for a one-time use during the 16th refueling outage in the third 10-year lSI interval. The third 10-year lSI interval began on began July 1, 2006, and is scheduled to end March 12. 2016. Granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will
J. Conway - 2 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.
All other ASME Code.Section XI. requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable. including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions regarding the SE. please contact James T. Polickoski at (301) 415-5430.
Sincerely,
~. ~ \ l,,,,' o.--t t /6, \;2~ --.):v Michael T. Markley. Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-323
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 RELATING TO RELIEF REQUEST NDE-RCS-SE-2R16 FOR THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL DIABLO CANYON POWER PLANT, UNIT 2 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323
1.0 INTRODUCTION
By letter dated August 12, 2010, Pacific Gas and Electric Company (PG&E, the licensee) submitted a request for relief NDE-RCS-SE-2R16 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102350309) from certain examination requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) at the Diablo Canyon Power Plant, Unit 2 (DCPP, Unit 2). Specifically, the licensee proposed using a root mean square error (RMSE) criterion for sizing flaws that is greater than ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds," (N-695). The licensee also requested relief from the N-695 exclusion for examinations through corrosion resistant clad (CRC). ASME Code Case N-695 is referenced in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," October 2007 (ADAMS Accession No. ML072070419). These requests are for the DCPP, Unit 2, 16th refueling outage (2R16),
scheduled to begin May 2011, which is in the third 1O-year year inservice inspection (lSI) interval.
2.0 REGULATORY EVALUATION
The lSI of ASME Code Class 1,2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The regulations in 10 CFR 50.55a(a)(3) state, 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.
Enclosure
- 2 Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," and to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-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 regulations in 10 CFR 50.55a(g)(4)(iv) states that 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 parqgraph 10 CFR 50.55a(b), subject to the limitations and modification listed in 10 CFR 50.55a(b) and subject to Commission approval. Portions of editions or addenda may be used, provided that all related requirements of the respective editions or addenda are met. The Code of record for the third 10-year lSI interval at DCPP, Unit 2, is the 2001 Edition with 2003 Addenda of the ASME Code with Appendix VIII of Section XI, 2001 Edition without Addenda.
3.0 TECHNICAL EVALUATION
FOR REQUEST 3.1 Affected Components Code Pipe Inside Category Diameter, Item Number Description Weld Number Inches R-A, R1.20 Loop 1, Outlet nozzle-to-safe-end WIB-RC-1-1 (SE) 29 Loop 1, Inlet nozzle-to-safe-end IB-RC-1-16 (SE) 27.5 Loop 2, Outlet nozzle-to-safe-end B-RC-2-1 (SE) 29 Loop 2, Inlet nozzle-to-safe-end B-RC-2-16 (SE) 27.5 Loop 3, Outlet nozzle-to-safe-end WIB-RC-3-1 (SE) 29 Loop 3, Inlet nozzle-to-safe-end WIB-RC-3-16 (SE) 27.5 Loop 4, Outlet nozzle-to-safe-end C-4-1 (SE) 29 Loop 4, Inlet nozzle-to-safe-end -RC-4-16 (SE) 27.5 3.2 Applicable Code Requirement The third 10-year lSI interval Code of record is the 2001 Edition through 2003 Addenda of the ASME Code,Section XI. For ultrasonic testing (UT) examinations, 10 CFR 50.55a(b)(2)(xv) requires the licensee to use the 2001 Edition ASME Code,Section XI, Appendix VIII, Supplement 10.
- 3 ASME Code Case N-695 is a Supplement 10 alternative that is endorsed by the NRC in RG 1,147, Revision 15, ASME Code Case N-695 states in the Scope, in part, that, This Case is not applicable to piping welds containing supplemental corrosion resistant clad (CRC) applied to mitigate intergranular stress corrosion cracking (IGSCC).
ASME Code Case N-695, paragraph 3.3(c), also states that, Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS [root mean square] error of the flaw depth measurements as compared to the true flaw depths, do not exceed 0.125 in. (3 mm).
3.3 Proposed Alternative (as stated by the licensee)
RMSE Error PG&E proposes to use approved [AS ME] Code Case N-695 with the demonstrated RMSE of 0.189 inches for 10 [inside diameter] examination of the nozzle-to-safe-end welds in lieu of the specified 0.125 inch RMSE. In the event an indication that requires sizing is detected, the 0.064 inch difference between the demonstrated RMSE and the required RMSE (0.189 inches minus 0.125 inches = 0.064 inches) will be added to the measured through-wall extent for comparison with the applicable acceptance criteria.
If advances in technology are realized and the contracted examination vendor demonstrates an improved RMSE for the Supplement 10 prior to the examinations, the difference of the improved RMSE over the 0.125 inch RMSE requirement, if any, will be added to the measured through-wall dimension of indications requiring sizing before comparison to the applicable acceptance criteria.
Code Case N-695 CRC Exclusion PG&E proposes to use vendor procedures, personnel and equipment qualified in accordance with the POI [Performance Demonstration Initiative] implementation of Appendix VIII, Supplement 10 as modified by the requirements of [ASME]
Code Case N-695 to examine the nozzle-to-safe-end dissimilar metal welds from the 10 through the protective clad layer.
PG&E's inspection vendor has conducted additional demonstration activities in order to validate the ability to detect, length size and depth size flaws through a clad layer and 10 weld inlays. Although not identical, the open test samples clad and weld layer thicknesses conservatively encompass the DCPP [Unit 2]
configuration. The test samples include flaws of various depth and lengths oriented in both the axial and circumferential directions. When examining the test specimens, the vendor used the same POI qualified (detection and length
- 4 sizing) procedure that will be employed in [refueling outage]2R16 examinations.
The results of these activities verify that the vendor has the capability to accurately detect, length and depth size the test sample flaws in each of the samples and configurations examined.
3.4 Licensee Basis for the Alternative (as stated by the licensee)
[ASME] Code Case N-695 contains an exclusion in the scope that states: 'This Case is not applicable to piping welds containing supplemental CRC applied to mitigate intergranular stress corrosion cracking [IGSCC)." The OCPP Unit 2 safe-end welds and safe-end forgings have a thin (0.073 inch to 0.125 inch in thickness) protective clad layer applied to the inside surface (exam surface) and outside surface of the dissimilar metal weld and the safe-end forging.
The OCPP [Unit 2] reactor coolant system (RCS) safe-end configuration with the protective 10 and outside diameter (00) clad layers applied to the dissimilar metal weld and safe-end forgings is unique to a small number of Westinghouse designed units. Suitable "blind" test samples are not available to support Supplement 10 qualification for this configuration. Removing the protective clad from either the 10 or 00 of the RCS safe-ends in order to create a configuration bounded by the performance demonstration initiative (POI) sample sets would result in extensive personnel exposure and potentially reduce the overall structural integrity of the component. General area dose rates in the vicinity of the subject welds (ex-core annulus area) averages 200 millirem per hour.
Considering that 00 machining to remove the overlay and achieve the required surface finish could exceed 20 man hours per nozzle, the total personnel exposure could surpass 32 rem [roentgen equivalent man]. 10 machining of these locations would remove the protective layer and any protection that it might afford to the underlying materials.
The OCPP Unit 2 reactor vessel was fabricated in the 1970 timeframe, prior to implementation of Appendix VIII qualification requirements. The distinctive
[OCPP] Unit 2 reactor coolant system safe-end weld configuration is not encompassed by the industry's PDI program used to implement ASME [Code, Section] XI, Appendix VIII requirements. Consequently, no inservice inspection vendor is qualified to examine the OCPP configuration. Additionally, the vendors are incapable of meeting the stringent 0.125 inch RMSE sizing accuracy requirement when examining from the 10 surface.
Compliance with the POI qualification program without alternative implementation would necessitate significant modification to the reactor coolant system safe end welds. Alterations such as this may result in reduced structural integrity of the reactor coolant pressure boundary. Even with such modifications, the vendor depth sizing accuracy issue would not be addressed.
-5 3.5 NRC Staff Evaluation The licensee's Code of record for the third 1O-year lSI interval is the 2001 Edition with 2003 Addenda. The ASME Code requires that dissimilar metal welds (OMWs) be examined using procedures, equipment, and personnel qualified to Section XI, Appendix VIII, Supplement 10.
The regulations in 10 CFR 50.55a(b)(2)(xv) require that ultrasonic examinations be performed using procedures, equipment, and personnel qualified to the 2001 Edition with no Addenda of Section XI, Appendix VIII, Supplement 10. However, the 2001 Edition does not provide criteria for examinations performed from the 10 of nozzles and piping. As an alternative to Supplement 10, the ASME Code developed Code Case N-695 for qualifications performed from either the 10 or 00 surfaces of OMWs. ASME Code Case N-695 is endorsed in RG 1.147, Revision 15 with no conditions.
ASME Code Case N-695 requires that the maximum error for flaw depth measurements, when compared to the true flaw depth, not exceed 0.125-inch RMSE. The U.S. nuclear power industry is using the POI program to implement the performance demonstration required by 10 CFR 50.55a. The RMSE is a statistical measurement with a screening criterion for separating skilled personnel and effective procedures from those that are less accomplished.
To date, no personnel and procedures have met the ASME Code Case N-695 depth-sizing qualification requirement, 0.125-inch RMSE maximum, for examinations performed from the 10.
The current POI program for qualifying procedures and personnel from the nozzle and pipe 10 is ineffective. The mockups used for the performance demonstrations are supposed-to-be representative of configurations common to the nuclear power industry. These mockups are fabricated to represent the extreme surface roughness (waviness) and pipe misalignment 10 conditions that may exist in RCS field welds. The surface roughness creates a gap beneath the probe and work piece that exceeds the 1/32-inch gap maximum recommended by the POI program. The industry's difficulty in meeting the RMSE requirement is associated with the POI as-built representative mockups and the surface condition that can be examined with the commonly used UT systems.
The OCPP, Unit 2 RCS nozzle-to-safe-end and safe-end-to-pipe configurations differ from the POI mockups in two respects. The 10 weld surfaces were machined smooth and the stainless steel safe-ends and welds were cladded on both 10 and 00 surfaces. In 2006, OCPP, Unit 2 made ID surface profilometry measurements of these welds. The profilometry measurements verified a smooth machined surface with sight contours at the start and stop of the cladding.
The POI program has no representative mockups of OCPP, Unit 2, cladded and smooth-surface configurations.
To show equivalency between examinations performed from cladded and unci added surfaces, the vendor performed a non-blind demonstration on a cladded (0.25-inch thick on one surface) weld block from another licensee containing two fatigue flaws and a second non-blind demonstration on a full scale OCPP, Unit 2, mockup with weld inlays in three of four 10 quadrants that ranged in thicknesses from 0.07 -inch to 1.00-inch. Each cladded quadrant contained four 10-connected flaws with one quadrant also containing an embed flaw. The flaws sized during the vendor's demonstration from cladded surfaces totaled 15. Collectively, the test block and the OCPP, Unit 2, mockup represented the geometrical aspects of the OCPP, Unit 2, nozzle-to-safe-end config urations.
- 6 For the non-blind equivalency demonstrations, the vendor used the same personnel and procedure that were used to achieve 0.189-inch RMSE from the blind POI performance demonstration. The 15 flaws from the equivalency demonstration provided a statistically significant accumulation of values for calculating depth and length RMSE values. The individually measured error values and calculated RMSE values from the equivalency demonstration were well within the Supplement 10 RMSE screening criteria. For comparison between examinations of cladded and uncladded configurations, the vendor sized four flaws in an unci added POI RCS nozzle-to-safe-end practice mockup that contained flaw sizes similar to those in the OCPP, Unit 2, mockup. The sizing errors for the comparison flaws from the POI practice mockup were within the sizing populations from the cladded mockups. The equivalency demonstration provides reasonable expectations that flaws from cladded surfaces would be detected and sized using the vendor's demonstrated procedure and personnel.
The licensee also proposed to perform an alternative eddy current testing (ET) surface examination. The ET would detect any surface breaking flaws and provide supporting information for the UT examinations.
In the absence of representative mockups of the OCPP, Unit 2, configuration for Supplement 10 qualifications, the licensee proposed applying the vendor's RMSE from the Electric Power Research Institute's (EPRl's) POI program for an approximation of the actual flaw depth to provide a reasonable level of depth-sizing capability. The licensee stated that the vendor's RMSE demonstrated on the current mockups in the POI program was 0.189-inch. The licensee proposed adding the depth-sizing difference between the demonstrated 0.189-inch RMSE and the ASME Code-required 0.125-inch RMSE to the measured value of any flaw detected during OMW examinations. The licensee stated that if advances in technology are realized and the contracted examination vendor demonstrates an improved RMSE for Supplement 10 prior to the examinations, the difference of the improved RMSE over the 0.125-inch RMSE, if any, will be added to the measured through-wall dimension of the indications.
The RMSE is a performance demonstration testing parameter used as a statistical measurement for screening individual skills and procedure capabilities. As a statistical measurement, the worst case error from a 0.125-inch RMSE is 0.395-inch for a performance demonstration test set with nine flaws measured precisely and the tenth flaw with maximum error. When calculating the worse-case error for a performance demonstration test set with 10 flaws and the proposed 0.189-inch RMSE, the worse-case error is 0.600-inch.
The licensee proposed to use 0.125-inch RMSE as an acceptable tolerance for subtracting from an individual's performance demonstrated RMSE. The application of RMSE as a tolerance for field applications has some inherent inaccuracies that normally exist between a performance demonstration environment (lax time constraints and ideal office environment) and field applications (outage constraints and field enVironment). Using the 0.125-inch tolerance adjustment to an individual's RMSE does not take into consideration the RMSE from a successful performance demonstration which is normally less than the Code-required maximum RMSE acceptance value.
-7 The test mockups used in the EPRI POI program were fabricated to replicate the bounding OMW surface conditions for pressurized-water reactor (PWR) cooling system piping. However, the EPRI POI program does not have cladded PWR cooling system mockups with less severe 10 surface roughness (waviness) for depth-sizing performance demonstrations. In the absence of cladded smooth-surface mockups, the licensee's vendor was unable to improve on its existing depth-sizing capabilities or achieve the ASME Code-required depth-sizing qualification.
Without representative mockups of the licensee's cladded PWR cooling system OMWs, the vendor is unable to qualify personnel and procedures to the ASME Code-requirement. The NRC and EPRI POI have been discussing the RMSE issue at semiannual public meetings with industry representatives.
Based on the above evaluation, the NRC staff concludes that compliance with the ASME Code Case N-695 required 0.125-inch RMSE, at this time, is impractical. Adding the difference between the performance demonstrate depth-sizing RMSE and the ASME Code Case N-695 required depth-sizing RMSE to a flaw size and applying the standards specified in ASME Code,Section XI, IWB-3500 to determine acceptability, provides reasonable assurance that structural integrity is being maintained for the subject OMWs.
4.0 Conclusion Based on the above review and evaluation, the NRC staff concluded that compliance with the ASME Code Case N-695 required 0.125-inch RMSE for depth sizing is impractical, and that the proposed alternative to add to the depth of a flaw the difference between 0.189-inch RMSE and the ASME Code-required value (0. 189-inch minus 0.125-inch = 0.064-inch) provides reasonable assurance of structural integrity of the OMWs that will be examined during the OCPP, Unit 2, refueling outage 2R 16. The NRC staff also concluded that the vendor-performed, cladded versus uncladded equivalency demonstrations provided reasonable expectation that flaws from cladded surfaces would be detected using the vendor's demonstrated procedures and personnel. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), relief is granted to OCPP, Unit 2, for a one-time use during the 16th refueling outage scheduled to begin May 2011, which is in the I
third 10-year lSI interval. The third 10-year lSI interval began July 1, 2006, and is scheduled to end March 12, 2016. The granting of 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, given the consideration of the burden upon the licensee that could result jf 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 request for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: O. Naujock Oate: March 29, 2011
J. Conway -2 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.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions regarding the SE, please contact James T. Polickoski at (301) 415-5430.
Sincerely, IRA by Balwant K. Singal fori Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket No. 50-323
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsNrrLAJBurkhardt Resource LPLIV R/F RidsNrrPMDiabloCanyon Resource RidsAcrsAcnw_MaiICTR Resource RidsOgcRp Resource RidsNrrDciCpnb Resource RidsRgn4MailCenter Resource RidsNrrDorlDpr Resource DNaujock, NRR/DCIICPNB RidsNrrDorlLpl4 Resource JTrapp, EDO RIV ADAMS Accession No. ML110600356 Memo OFF 4/PM . NRR/LPL4/PM NRR/LPL4/LA
..-----..---r~*-~---------*~-------*-----****--~!
NAM ki AWang JBurkhardt TLupold*
3/16/11 3/11/11 12/10/11 3129/11 OFFICIAL RECORD COpy