RNP-RA/04-0043, Response to Request for Additional Information for Relief Pertaining to Examination Coverage Less than Essentially 100 Percent (Relief Request No. 34)
| ML041070476 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 04/16/2004 |
| From: | Baucom C Carolina Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/04-0043 | |
| Download: ML041070476 (29) | |
Text
United States Nuclear Regulatory Commission Serial: RNP-RA/04-0043 Page 1 of 2 10 CFR 50.55a Serial: RNP-RA/04-0043 April 16, 2004 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST PERTAINING TO EXAMINATION COVERAGE LESS THAN ESSENTIALLY 100 PERCENT (RELIEF REQUEST NO. 34)
Ladies and Gentlemen:
In accordance with 10 CFR 50.55a(g)(6)(i), Relief Request No. 34 was submitted for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, by letter dated February 11, 2003.
A request for additional information (RAI) was received from the NRC by letter dated November 6, 2003. A response to that RAI was provided by letter dated December 30, 2003.
An additional RAI was received by letter dated March 5, 2004, with a requested response date of April 19, 2004.
The response to the additional RAI is provided in the attachments to this letter.
If you have any questions regarding this matter, please contact me.
Sincerely, Original signed by C. T. Baucom C. T. Baucom Supervisor - Licensing/Regulatory Programs CAC/cac
United States Nuclear Regulatory Commission Serial: RNP-RA/04-0043 Page 2 of 2 Attachments:
I. Response to Request for Additional Information for Relief Request Pertaining to Examination Coverage Less than Essentially 100 Percent (Relief Request No. 34)
II. Coverage Data for Welds 202/01, 202/02, and 204/A02 c: Mr. L. A. Reyes, NRC, Region II Mr. C. P. Patel, NRC, NRR NRC Resident Inspector
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 1 of 8 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RESPONSE TO REQUEST FOR ADDITONAL INFORMATION FOR RELIEF REQUEST PERTAINING TO EXAMINATION COVERAGE LESS THAN ESSENTIALLY 100 PERCENT (RELIEF REQUEST NO. 34)
The following response to the request for additional information (RAI) is provided:
Requested Information:
2.1 Omitted Welds Several welds included in original Relief Request 34 were omitted in the revision. In one case, it was stated that, upon further review of examination reports, the licensee determined that greater than 90-percent coverage had been obtained for certain welds; therefore, relief was not required. These welds are listed in Table 1 below. Please clarify that greater than 90-percent volumetric and surface, as applicable, examination coverage was obtained for the welds in Table 1 and how the decision to list these welds in the original request was reached.
Other welds appear to have been omitted in revised Relief Request 34 with no explanation. Please provide clarifying information as to why these welds listed in Table 2 were omitted.
Response
The welds listed in Table 1 of the current RAI identify three Category B-A and two Category B-D welds that were subsequently deleted in the response to the first request for additional information. The original relief request submittal was based on information identified from vendor supplied reports. After further review of the examination data for these welds, it was identified that the examination coverage was greater than 90%. This conclusion was based on information contained in the vendor final report. During preparation of the original relief request, the review of the vendor final report appeared to indicate that the coverage was 90% for these specific welds. After further review of the report during preparation of the previous RAI response, it was subsequently determined that the examination coverage was greater than 90%. Therefore, the initial relief request unnecessarily included these welds.
The welds listed in Table 2 of the current RAI identify three Category C-A and five Category C-F-1 welds. After further review, it has been determined that the three Category C-A welds were omitted inadvertently. The information requested by the previous RAI is provided for these welds in Attachment II to this letter.
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 2 of 8 Detail pertaining to the omission of the five Category C-F-1 welds is provided as follows:
Weld 231/30 - Safety Injection System Pipe to Elbow Weld This weld is part of the containment spray system, and is exempt from examination based on pipe wall thickness as described in ASME Section XI, IWC-2500. Twelve (12) containment spray system welds have been identified to be examined as part of the augmented examination of the containment spray system. Code credit is not needed for these augmented examinations. Therefore, this weld was deleted from the subsequent RAI response. These examinations are conducted on a best effort basis due to the thin pipe wall.
Welds 239/12, 239/13, 239/14, and 240/13 - Safety Injection System Pipe to Elbow Welds These welds were originally identified as a limited examination with 51.9%, 42.4%,
41.1%, and 86.32% examination coverage achieved, as identified on the examination coverage report. These coverages were subsequently revised to 100% during review by a Level III NDE examiner, because the contour plot revealed that the weld crown was less than three times the thickness and a 70° transducer was utilized to increase coverage.
Therefore, 100% coverage was credited as permitted by Performance Demonstration Initiative (PDI) letter, PDI Piping Generic Procedure Expansion for As Welded Condition, dated April 18, 2000.
Requested Information:
Finally, the welds listed in Table 3 below were omitted from the revised relief request with the following explanation:
During the review process performed on the subject welds, it was identified that the ISI Program description included steam generator nozzle safe end welds. After further review, it appears that the steam generators were provided with a stainless steel build-up on the hot leg and cold leg nozzles and that there is no weld in this location. Therefore, these welds are not included in this RAI response and have been deleted from the ISI Program description.
Please provide clarifying information on the weld configuration as the result of steam generator replacement and provide the dates for when the steam generators were replaced pertaining to these nozzle-to-piping welds. Please clarify that no new welds have been included in the ISI Program Plan. Also, provide any information related to limitations for examining these Category B-J welds or B-F welds.
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 3 of 8
Response
As stated in the previous RAI response, it has been determined that the steam generator nozzle welds each consist of one weld, not two welds as previously described in the ISI Program. These steam generator nozzle to reactor coolant loop elbow welds are configured with a buttered end preparation of the cast nozzle and are located between the nozzle and the cast elbow. Research into the initial examination requirements for these welds, as described in the original Technical Specifications (TS) for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, shows that these welds were called primary nozzle to safe-end welds and the TS identified six total welds. When this information was transferred to the ISI Program for the 2nd interval in the early 1980s, it appears that these locations were categorized as two welds, possibly by counting the butter as a separate weld or possibly due to the incorrect belief that a safe-end spool-piece existed at these locations.
The steam generator reactor coolant system (RCS) nozzle welds remain in the original construction configuration and were not affected by the steam generator replacement that was conducted at HBRSEP, Unit No. 2, during the Steam Generator Replacement Outage that ended on January 8, 1985. The steam generator replacements conducted at HBRSEP, Unit No. 2, were replacements of the tube bundle section of the three steam generators.
The steam generator RCS channel heads were cut just below the tube sheet and the new welds were placed at these locations. These new welds are Category B-B, Item No.
B2.40, Tubesheet-to-Head Weld, and are examined in accordance with ASME Section XI, Table IWB-2500-1 requirements.
Prior to deletion of the six reactor coolant system Category B-J steam generator nozzle welds from the ISI Program, the total weld count for the Third Ten-Year ISI Interval,
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 4 of 8 Category B-J, was 727 welds. The HBRSEP, Unit No. 2, extent of examination requirement for Category B-J welds is 25%, based on 10 CFR 50.55a(b)(2)(ii), which allows facilities with an application for construction permit docketed prior to July 1, 1978, to determine the extent of examination for Code Class 1 piping welds in accordance with Table IWB-2500 of Section XI of the ASME Code in the 1974 Edition and addenda through the Summer 1975 Addenda. The Category B-J total welds and completed examination count for the Third Ten-Year ISI Interval are summarized as follows:
Item Total 1st Period 2nd Period 3rd Period Number Completed Exams Completed Exams Completed Exams B9.11 184 20 20 20 B9.21 139 13 14 15 B9.31 7 0 0 2 B9.32 20 1 1 2 B9.40 377 32 29 40 727 66 64 79 With the deletion of the safe-end to elbow weld at each steam generator hot leg and cold leg (six welds deleted), the adjusted Category B-J totals are as follows:
Item Total 1st Period 2nd Period 3rd Period Number Completed Exams Completed Exams Completed Exams B9.11 178 18 18 18 B9.21 139 13 14 15 B9.31 7 0 0 2 B9.32 20 1 1 2 B9.40 377 32 29 40 721 64 62 77 The minimum number of welds required to be completed at the end of the Third Ten-Year ISI Interval for Category B-J was 181 (25% of 721). The revision to the program resulted in 203 examinations counted as completed, as opposed to the original count of 209. The examination count of 203 remains in excess of the Code-required minimum of 181.
As previously indicated, the Fourth Ten-Year ISI Interval Program has been revised to remove the six Category B-J welds that were listed as steam generator nozzle welds. The six remaining nozzle to elbow welds (previously identified as nozzle to safe-end welds) at the steam generator hot leg and cold leg nozzle locations are identified as Category B-F, Item Number B5.70. These six weld locations are scheduled for surface and volumetric examination, one set of steam generator welds each period during the interval. The additional welds in the ISI Program were introduced approximately 20 years ago.
Investigation of this situation has not revealed any other extra or missing welds. The extra welds did not cause any examinations to be conducted inappropriately.
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 5 of 8 Requested Information:
2.2 Request for Relief 34, Revision 1, Category B-D Examinations of Full-Penetration Welded Nozzles in Vessels Based upon the drawings and descriptions provided by the licensee for welds 101A/29, 101A/31, and 101A/33, it is not clear why only 10 percent of the subject nozzle weld can be examined when scanning transverse to the weld. In addition, the licensee does not state whether the subject weld was examined from the nozzle bore during the first period of the third inspection interval. Clearly explain why only 10 percent of the weld can be examined in the transverse direction and provide additional information to support a determination of reasonable assurance for continued structural integrity.
Response
The limitations relating to the transverse examinations of the hot leg reactor pressure vessel (RPV) nozzle to shell welds (101A/29, 101A/31, 101A/33) resulted in examination coverage that was reported as 10%. The reason for this low value is a combination of the physical obstruction presented by the nozzle integral extensions, the size of the transducers, and transducer mounting bracket. For these exams, a series of five (5) transducers are mounted on a crescent shaped bracket and are skewed to direct their respective beams towards the nozzle to shell weld in order to detect reflectors orientated transverse to the weld. However, due to the proximity of the weld to the outside diameter of the nozzle integral extension, the taper between the weld and the nozzle extension, and the narrow weld configuration (approximately 1 1/2 inches), the transducer assembly is physically restricted from the weld inner diameter surface. Also, because the nozzle forging is inserted into a curved vessel, the weld extends away from the inside diameter of the vessel at an angle approximately 15 degrees greater than the vessel radius, which further complicates the examination. As with all estimations of examination coverage, it should be noted that the coverage calculation is based on a theoretical point extending from the centerline of the transducers to the outer surface of the vessel at an angle normal to the vessel inside diameter. This calculation does not reflect the actual amount of ultrasound that has interrogated the weld due to beam-spread.
Requested Information:
2.3 Request for Relief 34, Revision 1, Category B-G-1 Examinations of Pressure-Retaining Bolting The licensee has requested relief from the 100-percent Code-required examination volume for reactor coolant pump stud No. 7. The licensee states that the lower 8 inches of the stud cannot be examined due to a taper in the stud design. Further, the licensee states that the limitation equates to only 0.36 percent of the overall examination volume for all studs in aggregate. We understand that this relief is no longer required. Please clarify your position.
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 6 of 8
Response
The original relief request identified stud No. 7 for the reactor coolant pump (RCP) C as a limited examination based on the NDE datasheet, which stated that the examination of the lower eight (8) inches was restricted due to taper. This datasheet was located in the final report for the refueling outage (RO) performed, which was RO-15. Subsequent to the examination of stud No. 7 on May 5, 1995, eight (8) studs were replaced (stud Nos. 1, 2, 3, 4, 5, 6, 23, and 24) due to a main flange leak. After additional visual and magnetic particle examinations, three (3) additional studs were replaced (studs Nos. 7, 8, and 21).
Preservice examinations were performed prior to installation of the studs. Based on this additional information, it has been determined that the ASME Section XI Code requirements have been met and that the examination of the installed components meet the applicable ASME Section XI Code requirements. Therefore, relief from the Code-required examination coverage is not being requested for RCP C stud No. 7.
Requested Information:
2.4 Confirm the end date for the H.B. Robinson third 10-year inspection interval was February 18, 2002.
Response
As stated in the Inservice Inspection Program for the Fourth Ten-Year ISI Interval letter, dated August 17, 2001, the Fourth Ten-Year ISI Interval for HBRSEP, Unit No. 2, began on February 19, 2002. This confirms that the Third Ten-Year ISI Interval ended on February 18, 2002.
Requested Information:
2.5 Request for Relief 34, Revision 1, Examination Categories B-F, B-J, C-F-1, and C-F-2 Pressure-Retaining Welds in Piping Clarify that 100 percent, with the exception noted for socket Weld 133/10, of the Code-required surface examinations were completed for all Class 1 and 2 dissimilar metal, austenitic, and ferritic piping welds included in Request for Relief 34. Briefly discuss any relevant indication, if observed, during the volumetric and surface examinations.
Response
The required surface examinations for the HBRSEP, Unit No. 2, Third Ten-Year ISI Interval for ASME Section XI, Categories B-F, B-J, C-F-1, and C-F-2, were performed as required. The volumetric examinations were also performed and did not reveal any rejectable indications.
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 7 of 8 Note that the ASME Section XI Code-required surface examinations were not performed on the reactor vessel nozzles, Category B-F (six welds) and B-J (six welds). This was allowed in accordance with Relief Request No. 32 for the Third Ten-Year ISI Interval.
These welds received an alternative VT-2 examination, which was approved by NRC letter dated June 6, 2001. The relief from surface examination requirements for these welds was needed due to physical limitations that restrict the ability to perform surface examinations on these welds.
During RO-15 in 1993, examination of reactor coolant pump seal injection system piping elbow socket weld 132/13 (Category B-J, Item No. B9.40) identified a one-half inch linear indication on the elbow. This indication exceeded the allowable acceptance criteria of the ASME Section XI Code. The indication was a shallow surface indication that was removed with a flapper wheel. A sample expansion was initiated and did not produce any further indications that exceeded the acceptance criteria of the ASME Section XI Code.
During RO-20 in 2001, high pressure safety injection system discharge piping elbow welds 239/18 and 239/19 were identified with indications. Both welds were examined in the as-welded condition. Examination of weld 239/18 identified three (3) straight linear indications 0.20 inch long separated by a 0.05 inch gap. The indication was not service induced and appeared to be a tooling mark. The indication was removed and the re-examination was satisfactory. The activity was documented on the NIS-2 Form and provided with the 90 day report for RO-20 in 2001, which was submitted to the NRC by letter dated August 10, 2001.
Examination of weld 239/19 identified a curvilinear indication 0.40 inch long. The indication was not service induced and appeared to be a lap formed during the welding process. The indication was removed and the re-examination was satisfactory. The activity was documented on the NIS-2 Form and provided with the 90 day report for RO-20 in 2001, which was submitted to the NRC by letter dated August 10, 2001.
As a result of the non-service induced indications identified on welds 239/18 and 239/19, a sample expansion was initiated in RO-20 as a conservative measure. The results of the sample expansion did not identify any additional indications. Welds 239/18 and 239/19 have been subsequently ground flush to allow for a complete PDI examination.
Requested Information:
2.6 Request for Relief 34, Revision 1, Examination Category C-C, Integral Attachments for Vessels, Piping, Pumps, and Valves Briefly discuss whether relevant indications have been detected on the subject integrally welded attachments, and describe other attachment welds that have received full Code examinations.
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0043 Page 8 of 8
Response
A review of the NDE data associated with ASME Code Category C-C integral attachments that were examined during the Third Ten-Year ISI Interval did not reveal any relevant indications. Limited examinations were conducted to the maximum extent practical.
Significant effort was required to complete the examination of the A residual heat removal (RHR) heat exchanger support welds due to as-found, as-welded conditions, which resulted in the liquid penetrant being trapped between weld passes producing a masking condition. After grinding, blending, and re-welding on the two (2) support welds, satisfactory examination results were achieved.
Typical integrally welded piping attachments that received ASME Section XI Code-required examinations consisted of piping lugs, ears, stanchions, pads, and saddles.
Components for which the Code-required examination coverage was achieved did not have physical limitations that prevented the Code-required surface area from being examined. The examinations for which Code-required examination coverage was not achieved, as identified in Relief Request No. 34, were due to physical access restrictions that did not allow access for liquid penetrant or magnetic particle examination.
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 1 of 19 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 COVERAGE DATA FOR WELDS 202/01, 202/02, AND 204/A02
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 2 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 3 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 4 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 5 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 6 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 7 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 8 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 9 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 10 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 11 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 12 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 13 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 14 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 15 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 16 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 17 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 18 of 19
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0043 Page 19 of 19