RS-09-142, Response to Request for Additional Information Regarding Relief Request I2R-50
| ML093000082 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 10/26/2009 |
| From: | Simpson P Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-09-142, TAC ME1038, TAC ME1039 | |
| Download: ML093000082 (11) | |
Text
Exelon Generation 4300 Winfield Road Warrenville, IL 60555 Nuclear RS 1 42 October 26, 2009 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457
Subject:
Response to Request for Additional Information Regarding Relief Request 12R-50
References:
- 1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S.
Nuclear Regulatory Commission, "Relief Request 12R-50, lnservice lnspection Program Relief Regarding Examination Coverage for Second 10-Year lnservice lnspection Interval," dated March 31, 2009
- 2. Letter from M. J. David (U. S. Nuclear Regulatory Commission) to C. G. Pardee (Exelon Generation Company, LLC), "Braidwood Station, Units 1 and 2 -
Request For Additional lnformation Related to Relief Request 12R-50 (TAC Nos. ME1038 and ME1039)," dated September 14,2009 In Reference 1, Exelon Generation Company, LLC (EGC), requested relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice lnspection of Nuclear Power Plant Components," for Braidwood Station Units 1 and 2, on the basis that compliance with the specified requirements is impractical due to plant design.
Specifically, 12R-50 requested relief for those ASME Section XI weld examinations performed during the second 10-year inservice inspection (ISI) interval where the inspection coverage achieved was less than or equal to 90%.
The relief request supports the second IS1 interval for both Units 1 and 2. The second IS1 interval for Unit 1 began on July 29, 1998, and concluded on July 28, 2009. The second IS1 interval for Unit 2 began October 17, 1998, and concluded on October 16, 2008.
In Reference 2, the NRC requested that EGC provide additional information in support of its review of Reference 1. In response to this request, Attachments 1 and 2 of this submittal provide the requested information.
October 26, 2009 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained in this submittal. Should you have any questions concerning this letter, please contact Ms. Lisa A. Schofield at (630) 657-281 5.
Res ectfully, R
\\
Patrick R. Simpson U
Manager - ~icensing Exelon Generation Company, LLC : Response to Request for Additional Information :
Supporting Information cc:
NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, Braidwood Station
ATTACHMENT 1 Response to Request for Additional Information In Reference 1, Exelon Generation Company, LLC (EGC), submitted relief request 12R-50, which requested relief for those ASME Section XI weld examinations performed during the second 10-year inservice inspection (ISI) interval where the inspection coverage achieved was less than or equal to 90%. In a letter dated September 14, 2009 (i.e., Reference 2), the NRC requested that EGC provide additional information related to relief request 12R-50 for Braidwood Station, Units 1 and 2. Attachments 1 and 2 provide the requested information.
- 1. Reactor Vessel Head-to-Flange Welds 1 RV-03-001 and 2RV-03-001 NRC Request a) The ASME Code,Section XI, requires essentially 100 percent volumetric coverage of the examination volume specified in Figure IWB-2500-5 for the reactor pressure vessel (RPV) head-to-flange welds. This examination volume includes the actual weld, as well as the adjacent base metal on either side of the weld extending to a distance of one-half the thickness of the vessel wall from the extremities of the weld crown. To obtain the total examination coverage, the licensee presents the calculated circumference in RR -2, "Weld 1 RV-03-001," sheet 11 of 19, and RR Attachment 2-2, "Weld 2RV-03-001," sheet 12 of 13, for the 1 RV-03-001 and 2RV-03-001 RPV head-to-flange circumference welds, respectively. This circumference value is used to determine the total weld metal volume, and subsequently the total inspection volume coverage. It is not clear to the NRC staff how this circumference value was calculated. Clarify the circumference calculations presented in the stated attachments.
Response
The figure in Attachment 1-2, sheet 1 1 of 19, and Attachment 2-2, sheet 12 of 13, of Reference 1 depicts a spherical segment representing the reactor pressure vessel (RPV) head-to-flange weld. Additional information about the weld is provided in of this submittal. In order to determine the circumference of the weld, the distance from the inside diameter weld centerline of the head-to-flange weld was calculated using the following formula for a spherical circular segment:
a = 2 R sin (112 8)
The nominal dimensions for the reactor vessel head provided in the coverage calculation sheets used the average vessel head radius for the "R" dimension. As shown in, the centerline of the RPV head-to-flange weld is located approximately 66 degrees (which is 112 8) below dead center of the reactor head (or 24 degrees above the RPV flange surface). The "a" dimension (denoted as "c" on the coverage calculation sketches) becomes the diameter, which multiplied by pi provides the nominal circumference of the RPV head-to-flange weld in inches.
Page 1 of 7
ATTACHMENT 1 Response to Request for Additional Information NRC Request b) The NRC staff requests that you discuss the extent to which these RPV head-to-flange welds were examined during the first IS1 interval, including the percentage of credible volumetric examination coverage that was achieved during these previous examinations.
Discuss any relevant conditions or indications that were found during these previous examinations.
Response
During the first IS1 interval, the achievable first IS1 interval volumetric coverage was calculated for welds 1 RV-03-001 and 2RV-03-001 in accordance with ASME Section XI 1983 Edition, summer 1983 Addenda requirements. Based on zero, 45, and 60 degree scans of the weld metal and base metal minus limitations associated with the three reactor head lifting lugs and not crediting scanning from the flange side of the weld, the examination coverage attainable was 53.25%.
There were no recordable indications documented on either unit's reactor head-to-flange weld during the first IS1 interval volumetric examinations.
- 2. Pressurizer Spray and Relief Nozzle-to-Vessel Welds 1 PZR-01 -N2.1 PZR-01 -N3, 2PZR-01 -N2 and 2PZR-01 -N3 NRC Request a) For the pressurizer spray and relief nozzle-to-vessel welds I PZR-01 -N2, 1 PZR-01 -N3, 2PZR-01 -N2 and 2PZR-01 -N3, discussed in RR Attachments 1-3, "Weld 1 PZR-01 -N2,"
1-4, "Weld 1 PZR-01 -N3," 2-3, "Weld 2PZR-01 -N2," and 2-4, "Weld 2PZR-01 -N3," were there any indications of the presence of unacceptable flaws or conditions found during the limited scope volumetric examination?
Response
There were no unacceptable flaws or conditions documented in any of the second IS1 interval limited scope volumetric examinations of welds 1 PZR-01 -N2, 1 PZR-01 -N3, 2PZR-01 -N2, or 2PZR-01 -N3.
NRC Request b) Discuss the extent to which these pressurizer spray and relief nozzle-to-vessel welds were examined during the first IS1 interval, including the percentage of credible volumetric examination coverage that was achieved during these previous examinations.
Discuss any relevant conditions that were found during these previous examinations.
Page 2 of 7
ATTACHMENT 1 Response to Request for Additional Information
Response
The first IS1 interval examinations of the pressurizer spray and relief nozzle-to-vessel welds were performed in accordance with the 1983 Edition, summer 1983 Addenda of ASME Section XI. The welds and weld heat affected zones were examined using zero degree longitudinal scans and single-sided axial and circumferential 45 and 60 degree shear wave scans performed from the vessel shell side of the welds. There were no percentages of volumetric examination coverage documented on the associated examination data sheets. There were no relevant indications documented for any of the first IS1 interval examinations of the pressurizer spray and relief nozzle-to-vessel welds.
- 3. Pipe-to-Valve Circumferential Weld 1 RC-17-13 NRC Request a) It appears to the NRC staff that there is a typographical error in RR Attachment 1-5, "Weld 1 RC-17-13." Section 3 references Figure 4-3, "Examination Volume for Thermal Fatigue Cracking of Sweepolets," from EPRl TR-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure." Clarify the RR to reference the correct figure, or confirm that the reference to Figure 4-3 is appropriate.
Response
In Reference I, the reference in Attachment 1-5, Section 3, to EPRl Report TR-112657, Figure 4-3, is a typographical error. The correct reference is Figure 4-2, "Examination Volume for Thermal Fatigue Cracking in Piping Welds NPS 4 or Larger."
In addition, the reference in Attachment 1-5, Section 3, to EPRl Report TR-112657, Figure 4-2, is also a typographical error. The correct reference is Figure 4-1, "Examination Volume for Thermal Fatigue Cracking in Piping Welds Less Than NPS 4."
These errors have been entered into the EGC corrective action program.
NRC Request b) For the pipe-to-valve circumferential weld 1 RC-17-13 discussed in RR Attachments 1-5, indicate whether the limited scope volumetric examination found any presence of indications during the examinations of the subject pipe-to valve circumferential welds.
Response
There were no recordable indications documented during the second IS1 interval limited scope volumetric examinations of Weld 1 RC-17-13.
Page 3 of 7
ATTACHMENT 1 Response to Request for Additional Information NRC Request c) The NRC staff requests that you discuss the extent to which the pipe-to-valve circumferential weld 1 RC-17-13 was examined during the first IS1 interval, including the percentage of credible volumetric examination coverage that was achieved during these previous examinations. Discuss any relevant conditions that were found during these previous examinations. Include clarification on the consideration of this weld as a "structural discontinuity" during the first IS1 inspection interval.
Response
The first IS1 interval volumetric examination of Weld 1 RC-17-13 was performed during the Braidwood Station Unit 1 spring 1994 refueling outage (i.e., A1 R04). The examination was a single-sided examination performed using a 45-degree shear wave.
No relevant indications were noted. The inside diameter (ID) geometry was observed and documented on the 1-112 vee path scan. The ID geometry was confirmed to be root convexity1concavity through a review of the original construction radiographs.
The term "structural discontinuity" is used in ASME Section XI Category C-F-1 for Class 2 pressure retaining austenitic welds and is defined as "pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, etc., conforming to ANSI B16.9), and pipe branch connections and fittings;"
however, the term was incorrectly referenced for this weld in Reference 1.
NRC Request d) RR Attachment 1-5, Section 4 states that the subject weld had been examined during the first IS1 inspection interval, and was reselected again in the second interval under the risk-informed IS1 program. Why was this weld selected for the risk-informed program?
Were there any other welds that could have been selected for which greater coverage could have been obtained? If so, why wasn't one of these welds selected?
Response
The subject weld is the classification break between the "RC" (reactor coolant) and "RY" (pressurizer) portions of the line. During development of the risk-informed IS1 program weld screening and selection, a number of single-sided welds downstream of the subject weld were recognized as single-sided examinations and removed from further consideration for the risk-informed IS! program, but the subject weld was overlooked as a single-sided examination. During the Braidwood Station Unit 1 spring 2006 refueling outage (i.e., A1 R12), six welds (including the subject weld) were selected for inspection.
The welds that were selected (all previously examined during the first IS1 interval) are located in a concentrated area, minimizing the amount of dose taken for craft support and inspection. The selection of the limited examination was due to an administrative error in the IS1 program. The subject weld has been flagged as a limited examination and alternate weld selections (pipe-to-elbow circumferential welds) on the line have been identified for consideration for the third IS1 interval.
The administrative error has been entered into the EGC corrective action program.
Page 4 of 7
ATTACHMENT 1 Response to Request for Additional Information
- 4. Steam Generator Auxiliary Feedwater Safe End-to-Nozzle Weld 1 SG-05-SGSE-02 NRC Request a) It appears to the NRC staff that there is a typographical error in RR Attachment 1-6, "Weld 1 SG-05-SGSE-02." Section 3 references Figure 4-3 from EPRI TR-112657.
Clarify the RR to reference the correct figure, or confirm that the reference to Figure 4-3 is appropriate.
Response
In Reference I, the reference in Attachment 1-6, Section 3, to EPRl Report TR-112657, Figure 4-3, is a typographical error. The correct reference is Figure 4-2, "Examination Volume for Thermal Fatigue Cracking in Piping Welds NPS 4 or Larger."
In addition, the reference in attachment 1-6, Section 3, to EPRI Report TR-112657, Figure 4-2, is also a typographical error. The correct reference is Figure 4-1, "Examination Volume for Thermal Fatigue Cracking in Piping Welds Less Than NPS 4."
These errors have been entered into the EGC corrective action program.
NRC Request b) It appears to the NRC staff that there is another typographical error in RR -6. Section 3 references ASME Code,Section XI, Table IWB-2500-1, in lieu of Table IWC-2500-1, for categorizing this weld. Clarify the RR to reference the correct table.
Response
The reference to Table IWB-2500-1 is a typographical error. Table IWC-2500-1 is the correct reference.
This error has been entered into the EGC corrective action program.
NRC Request c) Discuss the extent to which this steam generator auxiliary feedwater safe end-to-nozzle weld was examined during the first IS1 interval, including the percentage of credible volumetric examination coverage that was achieved during these previous examinations.
In addition, discuss any relevant conditions that were found during these previous examinations.
Response
The subject weld was installed during the Braidwood Station Unit 1 steam generator replacement project completed during the fall 1998 refueling outage (i.e., A1 R07). The A1 R07 refueling outage was the first refueling outage of the second IS1 interval, and the Page 5 of 7
ATTACHMENT 1 Response to Request for Additional Information subject weld did not exist during the first IS1 interval. Therefore, there are no previous inservice examinations applicable to discuss regarding this particular weld.
NRC Request d) Indicate if any other adjacent or similar welds were examined for this system. If any, indicate the percentage of credible surface examination coverage that was achieved. If less than essentially 100 percent coverage was achieved for any of these welds, provide supplemental information justifying why compliance with the ASME Code,Section XI requirements for essentially 100 percent volumetric examination coverage of these welds was impractical.
Response
The subject weld is a unique configuration (i.e., shop welded cap on an unused auxiliary feedwater nozzle on the replacement steam generators), using lnconel 690 materials.
Similar materials were used for the main feedwater (16") nozzle-to-safe end weld (ISG-05-SGSE-03), which was examined during the Braidwood Station Unit 1 fall 2001 refueling outage (i.e., A1 R09). Access from both sides of this weld was available, and the complete risk-informed IS1 coverage (1 00%) was achieved during this examination with no indications recorded.
NRC Request e) For the steam generator auxiliary feedwater safe end-to-nozzle weld 1 SG-05-SGSE-02, discussed in RR Attachment 1-6, indicate if any presence of indications were found in the subject auxiliary feedwater safe end-to-nozzle welds during the limited scope volumetric examinations.
Response
There were no recordable indications documented during the second IS1 interval limited scope volumetric examinations of weld 1 SG-05-SGSE-02.
- 5. Steam Generator Tube Sheet-to-Stub Barrel Weld 2SG-01-SGC-02 NRC Request a) For the steam generator tube sheet-to-stub barrel weld 2SG-01 -SGC-02, discussed in RR Attachment 2-5, "Weld 2SG-01 -SGC-02," indicate if any presence of indications were found in the subject steam generator tube sheet-to-stub barrel welds during the limited scope volumetric examinations.
Response
There were no recordable indications documented during the second IS1 interval limited scope volumetric examinations for weld 2SG-01 -SGC-02.
Page 6 of 7
ATTACHMENT 1 Response to Request for Additional lnformation NRC Request b) Discuss the extent to which this steam generator tube sheet-to-stub barrel weld was examined during the first IS1 interval, including the percentage of credible volumetric examination coverage that was achieved during these previous examinations. Discuss any relevant conditions that were found during these previous examinations.
Response
The first IS1 interval examinations of the steam generator tube sheet-to-stub barrel were performed in accordance with the 1983 Edition, summer 1983 Addenda of ASME Section XI. The welds and weld heat affected zones were examined using zero degree longitudinal scans and single-sided axial and circumferential 45 and 60 degree shear wave scans petformed from the vessel shell side of the welds. There is no examination coverage documented on the associated examination data sheets, but since the same angles were used for both the first and second IS1 intervals, it is reasonable to conclude that the examination coverage achieved would be similar, if not identical. There were no relevant indications identified for any of the first IS1 interval examinations. The tube sheet geometry was recorded under both the 45 and 60 degree shear wave scans from the stub barrel side of the weld.
References:
- 1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Relief Request 12R-50, lnservice lnspection Program Relief Regarding Examination Coverage for Second 10-Year lnservice lnspection Interval," dated March 31,2009
- 2. Letter from M. J. David (U. S. Nuclear Regulatory Commission) to C. G. Pardee (Exelon Generation Company, LLC), "Braidwood Station, Units 1 and 2 - Request For Additional lnformation Related to Relief Request 12R-50 (TAC Nos. ME1 038 and ME1 039)," dated September 14, 2009 Page 7 of 7
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