DCL-10-120, Response to NRC Letter Dated August 26, 2010, Request for Additional Information (Set 17) for License Renewal Application

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Response to NRC Letter Dated August 26, 2010, Request for Additional Information (Set 17) for License Renewal Application
ML102780501
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 09/24/2010
From: Becker J
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-10-120
Download: ML102780501 (21)
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Text

Pacific Gas and Electric Company James R.Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601 P 0. Box 56 Avila Beach, CA 93424 -

September 24, 2010 805.545.3462 Internal: 691.3462 Fax: 805.545.6445 PG&E Letter DCL-10-120 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20852 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Letter dated Auqust 26, 2010, Request for Additional Information (Set 17) for the Diablo Canyon License Renewal Application

Dear Commissioners and Staff:

By letter dated November 23, 2009, Pacific Gas and Electric Company (PG&E) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the license renewal application (LRA) and Applicant's Environmental Report - Operating License Renewal Stage.

By letter dated August 26, 2010, the NRC staff requested additional information needed to continue their review of the DCPP LRA.

PG&E's response to the request for additional information is included in Enclosure 1. LRA Amendment 14 resulting from the responses is included in the Enclosure 2 showing the changed pages with line-in/line-out annotations.

PG&E makes a commitment in revised LRA Table A4-1, License Renewal Commitments, shown in Enclosure 2.

If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 545-4-160.

I declare under penalty of perjury that the foregoing is true and correct.

Since ely, James R. Becker A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • San Onofre
  • South Texas Project e Wolf Creek

Document Control Desk PG&E Letter DCL-10-120 September 24, 2010 Page 2 pns/50338724 Enclosure cc: Diablo Distribution cc/enc: Elmo E. Collins, NRC Region IV Regional Administrator Nathanial Ferrer, NRC Project Manager, License Renewal Kimberly J. Green, NRC Project Manager, License Renewal Michael S. Peck, NRC Senior Resident Inspector Fred Lyon, NRC Project Manager, Office of Nuclear Reactor Regulation A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Caltaway

  • Comanche Peak ° Diablo Canyon
  • Palo Verde
  • San Onofre e South Texas Project e Wolf Creek

Enclosure 1 PG&E Letter DCL-1 0-120 Page 1 of 17 PG&E Response to NRC Letter dated August 26, 2010, Request for Additional Information (Set 17) for the Diablo Canyon License Renewal Application RAI 4.7.2-1 In license-renewal application (LRA) section 4.7.2, within the "Pressurizer" section, the applicantstates that the fatigue crack growth analyses were projected to the end of the period of extended operation and are therefore valid for the,period of extended operationin accordance with 10 CFR 54.21(c)(1)(i).

1. Discuss how the actual plant transient cycles are monitored to ensure that they are bounded by the number assumed in the fatigue crack growth analysis.
2. Discuss the transient cycles used in the crack growth analyses, including the number of cycles.

PG&E Response to RAI 4.7.2-1

1. The fatigue crack growth analyses associated with the Diablo Canyon Power Plant (DCPP) Unit 2 pressurizer structural weld overlays (SWOL) confirm that crack growth due to fatigue would remain within ASME Section X1, Appendix C, acceptable crack size criteria limits for 38 years after installation. The analyses are based on design basis numbers of transients. The SWOL were installed in 2008, therefore the analyses are valid through 2046, which encompasses the period of extended operation. Since the analyse's are valid through the end of period of extended operation, the TLAA for the SWOL fatigue crack growth is dispositioned in accordance with 10CFR54.21(c)(1)(i).

The actual plant transient cycles related to the SWOL fatigue crack growth analyses will be included in the existing plant transient monitoring program by January 31, 2011 to ensure that the actual plant transients do not exceed the SWOL fatigue analysis limits. See revised LRA Table A4-1 in Enclosure 2.

Enclosure 1 PG&E Letter DCL-10-120 Page 2 of 17

2. Transients used in the fatigue crack growth analysis are shown below with the number of cycles analyzed.

Unit 2 Pressurizer Unit 2 Pressurizer Unit 2 Pressurizer Transient Spray Nozzle Safety & Relief Surge Nozzle Nozzle Heatups/Cooldowns 250 250 310*

Unit Loading/ 41,950 -- 39,600 Unloading at 5 percent/min Red. Temp. Return to 4,470 Power Large Step Load 250 250 200 Decrease w/Steam Dump _

10 percent Step Load 2,500 each 2,500 4,000 Increase/Decrease Boron Equalization 32,000 -- 32,000 Loss of Load 100 100 80 Loss of Power 50 50 40 Loss of Flow 100 .100 220 Reactor Trip 500 500 400 Inadvertent Auxiliary 12 12 24 Spray Operation Basis 400 400 400 Earthquake Load Cycles Turbine Roll 10 10 20

  • Combines heatup and cooldowns with 60 leak test transients.

Two transients used in the fatigue crack growth analysis have been deemed nonsignificant: (1) Reduced temperature return to power, and (2) Boron equalization per the Westinghouse system standard. These transients are associated with load following. The current operating strategy for the DCPP units is continuous base-load power generation. Therefore, the actual number of reduced temperature return to power and boron equalization occurrences is expected to be a small fraction of the cycles assumed in the fatigue analyses.

Enclosure 1 PG&E Letter DCL-10-120 Page 3 of 17 RAI 4.7.2-2 In LRA section 4.7.2, within the "Pressurizer" section, theapplicantstates that "[n]o base-metal corrosionanalyses exist for the pressurizers,since no half-nozzle or similar repairs have exposed the base metal to reactorcoolant." The applicant also states that

"[tihe Unit I pressurizerand its nozzles and safe ends contain no Alloy 600 or Alloy 82/182 weld material." The above statements are not clear regardingwhether the half nozzle method was used in repairingheatersleeves in the pressurizerin both units.

1. Foreach unit, list all the pressurizernozzles (e.g., pressurizersafety valve nozzle and heatersleeve nozzle). Identify the materials used to fabricate the nozzles. If a nozzle is welded to a safe end, identify the materialof the safe end.
2. Discuss whether a fatigue crack growth calculation was performed for the remnant Alloy 82/182 welds. If so, discuss how the transient cycles used in the fatigue crack growth calculation are monitored to ensure they bound the actual plant cycles. If no fatigue crack growth calculation was performed,justify the structuralintegrity of the pressurizershell.
3. Discuss any flaws that remained in service in the heater sleeves and in the attachment welds in both units. If so, discuss how these flaws are monitored and evaluated for the period of extended operation.

PG&E Response to RAI 4.7.2-2

1. Pressurizer nozzles and materials of fabrication are provided below. If a nozzle is welded to a safe end, the material of the safe end is also identified.

Component Unit I Unit 2 Surge Nozzle Nozzle - SA-216 WCC Nozzle - SA-508 Cl. 2 and Safe End Material Safe End - SA-182 Type 316 Safe End - SA-182 Type 316L Spray Nozzle Nozzle - SA-216 WCC Nozzle - SA-508 Cl. 2 and Safe End Material Safe End - SA-182 Type 316 Safe End - SA-182 Type 316L Safety and Nozzle - SA-216 WCC Nozzle - SA-508 Cl. 2 Relief Nozzle and Safe End Safe End - SA-182 Type 316 Safe End -,SA-182 Type 316L Material Instrument SA-213 Type 316 SA-213 Type 316 Tube Heater Well SA-213 Type 316 SA-213 Type 316

Enclosure 1 PG&E Letter DCL-10-120 Page 4 of 17

2. Westinghouse performed an assessment of primary water stress corrosion cracking susceptibility for Alloy 82/182 welds in Diablo Canyon Power Plant (DCPP) Units 1 and 2 as discussed in License Renewal Application (LRA) Section 4.7.2. The only pressurizer repair or mitigation work that has been completed at DCPP is the Unit 2 pressurizer structural weld overlays as discussed in LRA Section 4.7.2 and in PG&E's response to Request for Additional Information 4.7.2-1.
3. No flaws have been identified in DCPP Units 1 or 2 pressurizers.

Enclosure 1 PG&E Letter DCL-10-120 Page 5 of 17 RAI 4.7.2-3 Discuss whether reactorvessel internals contain any nickel-basedAlloy 600 components or nickel-basedAlloy 82/182 welds. If so, discuss how these components are monitored for primary water stress corrosioncracking.

PG&E Response to RAI 4.7.2-3 The Diablo Canyon Power Plant reactor vessel internals do not contain any Alloy 600 components or nickel-based Alloy 82/182 welds.

Enclosure 1 PG&E Letter DCL-10-120 Page 6 of 17 RAI 4.7.2-4 In LRA section 4.7.2, within the "Steam Generators"section, the applicantstates that

"[rieplacementsteam generatorscontain no Alloy 600 components or Alloy 82/182 welds."

1. Identify the materialspecification of the welds that join the replacementsteam generatornozzles to the piping.
2. Identify the materialspecification of the safe ends that are weldedto the steam generatornozzles.

PG&E Response to RAI 4.7.2-4 The table below provides the requested materials associated with the Diablo Canyon Power Plant Units 1 and 2 replacement steam generators (RSGs).

Material specification of the Material specification of the RSG Nozzle welds that join the RSG safe ends that are welded to nozzles to the piping the RSG nozzles Primary Nozzles ER316L SA-336 Class F316LN Feedwater Nozzles ER70S-6 & E7018 SFA-5.18 Class ER70S-X Steam Nozzles ER70S-6 & E7018 SA-508 Grade 1A

Enclosure 1 PG&E Letter DCL-10-120 Page 7 of 17 RAI 4.7.5-1 In LRA Section 4.7.5, within the "Unit 2 RHR Piping Weld RB- 119-11"section, the applicant states that "[t]heDCPPlicensing basis assumes 250 heatups and 250 cooldowns for a 50 year plant life."

1.' Discuss why only heatup and shutdown cycles are applied for flaw evaluation of weld RB- 119-11 in the June 6, 2006 letter but other transient cycles such as seismic, temperature, and pressure were not mentioned in the flaw evaluation for weld RB-119-11.

2. It is not clear in LRA Section 4.7.5 or in the flaw evaluation that the cycles used in the flaw evaluation for weld RB- 119-11 bounds the accumulated transient cycles at the end of 60 years. LRA section 4.7.5 states that "[t]heservice life' for Weld RB- 119-11 is based on operating for 40 years from the date the flaw was identified, i.e. until 2046, during which the flaw would experience 500 startup-shutdown cycles. Thus, the evaluation encompassed a 60-year plant life and the analysis will be valid beyond the 2045 end date of the period of extended operation for Unit 2." The above statements do not provide a clearreasoningas to how the flaw evaluation for 40 years encompasses 60 years of plant life.

Clarify how the flaw evaluation encompassed a 60 year plant life in terms of cycle counting (e.g., are the 500 startup and shutdown cycles bound the actual plant cycles at the end of 60 years?).

3. Discuss how you ensure that transientcycles used in the flaw evaluation for the Unit 2 residualheat removal (RHR) piping weld RB-1 19-11 do not exceed the actual operatingcycles at the end of 60 years without the enhanced fatigue management program.
4. (a) Provide the materialspecification of weld RB- 119-11 (e.g., E308L or Alloy 82/182). (b) Discuss whether the indication in weld RB-1 19-11 is surface-connected or embedded. (c) Discuss the degradationmechanism of the indication.' (d) If the weld is fabricated with Alloy,82/182 metal or if the flaw is embedded in the pipe/weld wall thickness, discuss any mitigation measures applied to the flaw in Weld RB-1 19-11.
5. Discuss whether weld RB-119-11 will be examined in the future ASME 10-year inservice inspection (ISI) intervals. If not, provide justifications.

II

Enclosure 1 PG&E Letter DCL-10-120 Page 8 of 17 PG&E Response to RAI 4.7.5-1

1. Only heatup and shutdown cycles were discussed in the License Renewal Application (LRA) for the Diablo Canyon Power Plant (DCPP) Unit 2 residual heat removal (RHR) piping weld RB-119-11 flaw evaluation because the flaw evaluation only used heatup and shutdown cycles. Maximum stresses (not cycles) due to pressure, deadweight, seismic loadings, and thermal expansion were also used in the evaluation.

The Unit 2 RHR piping weld RB-1 19-11 flaw evaluation was submitted to the NRC in PG&E Letter DCL-06-069, "Residual Heat Removal Weld RB-1 19 Flaw Analytical Evaluation Results," dated June 6, 2006. The flaw evaluation was performed based on the guidelines of ASME Code,Section XI, IWB-3640, to calculate the allowable flaw size for the RHR pipe weld, specifically using the procedures and acceptance criteria of IWB-3641.

2. As shown in LRA Table 4.3-2, the number of heatup and cooldown cycles that DCPP projects for 60 years of operation (based on actual plant operating history) is 65 and 63 for Unit 2, respectively. This is less than the 500 heatup and cooldown cycles that were used in the Unit 2 RHR piping weld RB-1 19-11 flaw evaluation.

Thus, the flaw evaluation cycles are bounded by projected actual plant cycles at the end of 60 years.

Additionally, as shown in LRA Table 4.3-2, the transient cycles used in the flaw evaluation for the Unit 2 RHR piping weld RB-1 19-11 (plant heatup and cooldown cycles) are monitored by the Metal Fatigue of Reactor Coolant Pressure Boundary Program, as summarized in LRA Section B3.1. The Metal Fatigue of Reactor Coolant Pressure Boundary Program will ensure that transient cycles used in the flaw evaluation are not exceeded by the actual operating cycles.

3. Since the Unit 2 RHR piping weld RB-1 19-11 flaw evaluation states that it is valid through October 2046, the time limited aging analysis (TLAA) has been dispositioned in accordance with 10 CFR 54.21 (c)(11)(i). Additionally, as described in part 2 of this response, it has been shown (based on actual plant operating history) that the flaw evaluation cycles are bounded by projected actual plant cycles at the end of 60 years.

As required by the Metal Fatigue of Reactor Coolant Pressure Boundary Program (as summarized in LRA Section B3.1), if DCPP reaches one of the cycle count action limits, acceptable corrective actions are implemented.

Enclosure 1 PG&E Letter DCL-10-120 Page 9 of 17

4. (a) The weld filler material that is used for RB-1 19-11 is ER308.

(b) The indication in weld RB-119-11 is embedded.

(c) The flaw was characterized as a lack of fusion from original fabrication and was not service induced.

(d) No mitigation measures were applied to the flaw in Weld RB-1 19-11.

5. As required by IWA-2420 of ASME Code, Section Xl, one successive examination was completed for the Unit 2 RHR piping weld RB-1 19-11 flaw. The ultrasonic examination concluded that there were no apparent changes in the indication and that the results were satisfactory. As required by the ASME Code, Section Xl, Unit 2 RHR piping weld RB-1 19-11 will be examined in the future ASME 10-year in service inspection intervals.

Enclosure 1 PG&E Letter DCL-10-120 Page 10 of 17 RAI 4.7.5-2 LRA section 4.7.5 discusses the flaw evaluation of an indication detected in weld WIC-95 of the RHR injection line 985 to hot legs 1 and 2 as shown in Pacific Gas and Electric Company (PG&E)Letter DCL-97-086 dated May 7, 1997. LRA Section 4.7.5 states further that "[t]herehave been no occurrences of a DE, DDE, or Hosgri seismic event at Diablo Canyon Power Plant (DCPP)during the first 20 plus years of operation.

Therefore, the seismic cycles in the Unit I RHR Weld WIC-95 fatigue crack growth evaluation for the 50-year design basis number of DE, DDE, and Hosgri events are sufficient to the end of the period of extended operation."

1. LRA section 4.7.5 states that "[tihe number of seismic cycles used in the analysis

[flaw evaluation]is consistent with the DCPP50-year design basis described in FSAR Table 5.2-4..." FinalSafety Analysis Report (FSAR) Table 5.2-4 specifies one cycle for the Hosgri earthquake, 20 cycles for the design earthquake (DE),

and I cycle for the double design earthquake (DDE). In the flaw evaluation for weld WIC-95 in the applicant'sletter dated May 7, 1997, none of these seismic cycles were discussed. The applicant'sflaw evaluation discussed only "400 cycles of future loading for the governing pipe stress load case" Clarify whether the seismic cycles were included in the flaw evaluation of the indication at weld WIC-95.

2. FSAR Table 5.2-4 provides several transientsthat have more occurrences/cycles than 400 cycles used in the flaw evaluation for weld WIC-95.

Forexample, Unit loading and unloading at 5% of full power has 18,300 occurrences (cycles), hot standby operation/feedwatercycling has 18,300 occurrences. (a) Identify the transientsthat are included in the 400 cycles. (b)

Provide basis for those transientsshown in Table 5.2-4 but were not included in the flaw evaluation for weld WIC-95.

3. FSAR Table 5.2-4 specifies 250 occurrences for reactorcoolant system heatup and cooldown transients. The total cycles for heatup and shutdown transients would be 500. However, the flaw evaluation used only 400 cycles. The staff notes that 500 cycles were used in the flaw evaluation of the indicationin weld RB-119-11. The cycles in FSAR Table 5.2-4 are for the design life of the plant which presumably is 50 years. It appears that the 400 cycles used in the flaw evaluation for weld WIC-95 are for 50 years, not 60 years, of plant operation.

LRA section 4.7.5 states that the seismic cycles in the weld WIC-95 fatigue crack growth evaluation for the 50-year design basis number of DE, DDE, and Hosgri events are sufficient to the end of the period of extended operation. Clarify whether (a) the seismic cycles in the flaw evaluation in the May 7, 1997 letter, are sufficient to cover the seismic cycles at the end of extended operation, (b) the 400 cycles cover all the transientcycles at the end of extended operation, and (c) why a total of 500 cycles for heatup and cooldown were not used.

Enclosure 1 PG&E Letter DCL-10-120 Page 11 of 17

4. (a) Provide the pipe diameterand wall thickness at weld WIC-95 of the Unit 1 RHR injection line 985 where an indication was detected in refueling outage 9.

(b) In the flaw evaluation dated May 7, 1997, the applicantstated that it will re-examine the indication in weld WIC-95 in refueling outage IR1O. Discuss the inspection result of weld WIC-95 during refueling outage IR10. Confirm that the indication was detected in 1997 and was re-examined in 1999. (c) Provide the material specification of weld WIC-95 (e.g., Alloy 82/182 weld or E308L). (d)

Discuss whether the subject indication is surface-connected or embedded. (e)

Discuss the degradationmechanism of the indication. (f) Discuss the orientation of the indication (i.e., a circumferentialor an axial indication). (g) Provide operatingtemperature and pressure of the subject pipe line at weld WIC-95.

5. Discuss whether weld WIC-95 will be examined in the future ASME 10-year ISI inspection intervals. If not, provide justifications.
6. It is not clearto the staff that the applicanthas demonstrated that the cycles used in the flaw evaluation for weld WIC-95 bounds the cycles at the end of 60 years. Discuss how you ensure that transientcycles used in the flaw evaluation for the RHR piping weld WIC-95 do not exceed the actual operatingcycles.

PG&E Response to RAI 4.7.5-2

1. Cycles for the design earthquake were included in the Unit 1 residual heat removal (RHR) Weld WIC-95 flaw evaluation. As stated in PG&E Letter DCL-97-086, "Inservice Inspection Evaluation Analysis of Flaw Indication for Weld WIC-95 (Reference A0430829)," dated May 7, 1997, '400 cycles of future loading for the governing pipe stress load case" were assumed. The flaw evaluation further clarifies that these "400 cycles of future loading" are seismic cycles. This is consistent with Final Safety Analysis Report (FSAR) Table 5.2-4, which states that the 50-year design basis for design earthquakes is 20 events, with 20 cycles per event (a total of 400 cycles).
2. (a) As stated in part I of this response, the "400 cycles of future loading" are seismic cycles.

(b) The Unit 1 RHR Weld WIC-95 flaw evaluation was performed based on the guidelines of ASME Code,Section XI, IWB-3640, to calculate the allowable flaw size for the RHR weld.

RHR injection line 985 to hot legs 1 and 2 only operates during plant refueling (i.e., during heatups and cooldowns). When not in a plant refueling mode, the RHR injection line is not in service. Thus, those additional transients listed in FSAR Table 5.2-4 have no significant impact on the line and do not contribute any thermal cycles. The Unit 1 RHR Weld WIC-95 flaw evaluation states that the seismic events, plus pressure and deadload, envelops the thermal stress

Enclosure 1 PG&E Letter DCL-10-120 Page 12 of 17 both in magnitude and number of cycles. Additionally, thermal and seismic stresses are not combined per ANSI B31.1 code.

3. (a) The 400 seismic cycles used in the flaw evaluation are adequate for the period of extended operation because no seismic cycles have occurred at Diablo Canyon Power Plant (DCPP) since operation began. As with all other transients, seismic cycles are projected to 60 years of operation by using the actual plant operating history and projecting it to 60 years. As shown in License Renewal Application (LRA), Table 4.3-2, the projected number of design earthquakes (and thus the number of seismic cycles) is less than the 400 cycles used in the flaw evaluation.

(b) As stated in part 2(a) of this response, the 400 cycles used in the flaw evaluation are seismic cycles. The flaw evaluation did not address other transient cycles because, as stated in Request for Additional Information Response 4.7.5-2, part 2(b), those additional transients listed in FSAR Table 5.2-4 have no significant impact on the line and do not contribute any thermal cycles.

(c) Heatup and cooldown cycles were not included in the Unit 1 RHR Weld WIC-95 flaw evaluation. Rather, the flaw evaluation used 400 future loading cycles because seismic events, plus pressure and deadload, enveloped the thermal stress (which would be associated with heatups and cooldowns) both in magnitude and number of cycles.

4, (a) The pipe diameter and wall thickness at weld WIC-95 of the Unit 1 RHR injection line 985 where an indication was detected was 12.750 inches outside diameter and 0.410 inches, respectively.

(b) As required by IWA-2420 of ASME Code,Section XI, one successive examination was completed for the Unit 1 RHR Weld WIC-95 flaw in October 1999. The ultrasonic examination concluded that there were no apparent changes in the indication and that the results were satisfactory.

(c) The material specification of Weld WIC-95 is ER308.

(d) As stated in PG&E Letter DCL-97-086, the subject indication is inside diameter connected.

(e) The flaw was characterized as construction-related flaw and was not service induced.

(f) The orientation of the indication is circumferential.

Enclosure 1 PG&E Letter DCL-10-120 Page 13 of 17 (g) The maximum operating temperature and pressure of the subject line at weld WIC-95 are 350°F and 700 psig, respectively.

5. As required by the ASME Code,Section XI, Weld WIC-95 will be examined in the future ASME 10-year in service inspection intervals.
6. Since the Unit 1 RHR Weld WIC-95 flaw evaluation shows that the flaw is valid after 400 seismic cycles, the time limited aging analysis has been dispositioned in accordance with 10 CFR 54.21(c)(1)(i). Additionally, as described in LRA Section 4.7.5, it has been shown (based on actual plant operating history) that the flaw evaluation seismic cycles are bounded by projected actual plant cycles at the end of 60 years.

.As required by the Metal Fatigue of Reactor Coolant Pressure Boundary Program, (as summarized in LRA Section B3.1), if DCPP reaches one of the cycle count action limits (such as for seismic cycles), acceptable corrective actions are implemented.

Enclosure 1 PG&E Letter DCL-10-120 Page 14 of 17 RAI 4.7.5-3 LRA Section 4.7.5 discussed the indication detected in Unit 2 Auxiliary feedwaterpiping line 567. The applicantsubmitted a flaw evaluation in PG&E letter DCL-99-136, dated October 22, 1999.

1. In the flaw evaluation for piping line 567, the applicantstated that it will re-examine the indication during the Unit 2 tenth refueling outage (2R10). Discuss the inspection results of the re-examination.
2. The applicant stated in the flaw evaluation that the indication is believed to be a fabrication defect (a lap in the pipe). Confirm that the indication is embedded in the pipe wall. As stated in the flaw evaluation, the flaw was characterizedas 0. 1 inch deep (approximately46 percent through wall) and 12 feet in length.

Describe in detail how the indication is modeled in the flaw growth calculation.

3. The flaw evaluation dated October22, 1999 states that the 250 cycles of future seismic and thermal loading correspondingto the remaining plant life. In LRA Section 4.7.5, the applicantstated that the assumed transients are consistent with or bounded by the 50 year design basis described in FSAR Table 5.2-4. It is not clear to the staff that 250 cycles used in the flaw evaluation bound the cycles in Table 5.2-4 in FSAR. Identify the transients that are included in the 250 cycles. Discuss in detail how 250 cycles in the flaw evaluation bound the cycles in the licensing basis.
4. Discuss whether the indication in Unit 2 Auxiliary feedwaterpiping line 567 will be examined in the future ASME 10-year ISI inspection intervals. If not, provide justification.

PG&E Response to RAI 4.7.5-3

1. One successive examination was'completed for the Unit 2 auxiliary feedwater piping line 567. The ultrasonic examination concluded that there were no apparent changes in the indication and that the results were satisfactory.
2. The indication in the Unit 2 auxiliary feedwater piping line 567 was surface-connected, not embedded.

Because the piping material is carbon steel with stresses in the elastic range, the associated flaw evaluation used linear elastic fracture mechanics to evaluate the flaw growth. This approach is conservative since the carbon steel material has some ductility. The methodology is similar to ASME Section XI Appendix A, except that the Appendix A crack growth relations are based on a flat plate, while the analysis is performed for cylindrical geometry and is thus more accurate for a pipe.

Enclosure 1 PG&E Letter DCL-10-120 Page 15 of 17 The flaw model used was a longitudinal crack in a cylinder with t/R=0.2 (i.e., the ratio of pipe thickness to pipe mean radius). All of the stresses were conservatively applied as membrane stresses. Using the crack growth law for ferritic.steel in an air environment and the material fracture toughness of carbon steel, the crack growth was determined for the given number of cycles. It was determined that the growth in the flaw was below the critical flaw size.

3. The Unit 2 auxiliary feedwater line 567 flaw evaluation considered 250 Hosgri seismic loads (5 seismic events with 50 cycles per event). This is more conservative than the licensing basis described in Final Safety Analysis Report, Table 5.2-4, because it is based on 5 Hosgri events while the licensing basis only anticipates 1 event.
4. Diablo Canyon Power Plant (DCPP) evaluated the Class 3 Unit 2 auxiliary feedwater line 567 flaw to Class 1 requirements since the 1989 ASME Code then in effect did not have Class 3 acceptance criteria. Although there is no applicable requirement, DCPP committed to perform one successive exam, which yielded satisfactory results. There are no plans to conduct any further inspections on the Unit 2 auxiliary feedwater line 567 because; (1) it is not required, (2) the flaw is a fabrication defect and is not service-related, and (3) a follow-up examination showed there was no change in the flaw.

Enclosure 1 PG&E Letter DCL-10-120 Page 16 of 17 RAI B2.1.39-1 In LRA Section B2.1.39; the applicantstates that the Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program will be implemented as part of the A SME Code, Section Xl ISI program and will be completed within the 10-year inspection interval before the period of extended operation.

1. The NRC staff notes that ultrasonictesting (UT) has not yet been qualified to examine CASS material via the ASME Code, Section Xl, Appendix VIII. Discuss how components fabricated with CASS materialare inspected under the current licensing basis. Discuss whether the current inspection practices (methods, frequencies and acceptance criteria) will be applied in the future CASS aging managementprogram (AMP).
2. In light of the limitation of UT of CASS material,discuss how volumetric examination of CASS components will be accomplished during the period of extended operation. Specifically, clarify whether the qualified UT will only be used in the CASS AMP, if a qualified UT method becomes available.

PG&E Response to RAI B2.1.39-1

1. Components fabricated with cast austenitic stainless steel (CASS) material that are in scope of Aging Management Program (AMP) B2.1.39, reactor coolant loop elbow fittings, are currently pressure tested every refueling outage per the current ASME Section XI Code edition in effect. Current inspection practices will continue during the period of extended operation as required per the ASME Code editions in effect during the period of extended operation. In addition, as indicated in License Renewal Application (LRA) Section B2.1.39, for in-scope CASS components that are determined to be susceptible to the aging effect of thermal embrittlement, aging management would be accomplished through a qualified volumetric examination, provided one is demonstrated to be adequate for CASS inspection in accordance with criteria identified in ASME Section XI, Appendix VIII, or a component-specific flaw tolerance evaluation will be performed. Additional inspection or evaluations to demonstrate that the material has adequate fracture toughness will not be required for components that have been determined to not be susceptible to thermal aging embrittlement.
2. For the CASS AMP, DCPP will either; (1) use a qualified ultrasonic testing (UT) method for enhanced volumetric examination, if one becomes available, or (2) perform a component-specific flaw tolerance evaluation. As indicated in LRA Section B2.1.39, this AMP is a new program and if a viable volumetric examination method is developed, it will be implemented as part of the Section XI In Service Inspection Program. The qualified UT method will be demonstrated to be adequate for CASS inspection in accordance with criteria specified in ASME Section XI, Appendix VIII.

Enclosure 1 PG&E Letter DCL-10-120 Page 17 of 17 RAI B2.1.39-2 (1) Discuss whether Diablo Canyon units I and 2 have implemented the risk-informed ISI program. (2) If yes, discuss how the CASS components will be inspected under the risk-informed ISI program considering the requirementsof the CASS aging management program (e.g., whether the CASS AMP will increase the inspection frequency of the CASS components in the risk-informed ISI program and whether thermal aging embrittlement will be a degradationmechanism considered in the risk-informed ISI program).

PG&E Response to RAI B2.1.39-2

1. -For the current 10-year in service inspection (ISI) interval, Diablo Canyon Power Plant, Units 1 and 2, have implemented the risk-informed ISI Program for piping welds.
2. Current inspection practices (pressure tests) will continue during the period of extended operation as required per the ASME Code editions in effect during the period of extended operation. In addition, regardless of whether the ISI Program for the period of extended operation is risk informed, as indicated in License Renewal Application, Section B2.1.39, for in-scope cast austenitic stainless steel components that are determined to be susceptible to the aging effect of thermal embrittlement, aging management will be accomplished through a qualified volumetric examination, if one becomes available, once every 10 years. Alternatively, a component-specific flaw tolerance evaluation will be performed.

PG&E Letter DCL-10-120 Page 1 of 2 LRA Amendment 14 LRA Section RAI Table A4-1 RAI 4.7.2-1 TABLE A4-1 PG&E Letter DCL-10-120 LICENSE RENEWAL COMMITMENTS Page 2 of 2 Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 38 The actual plant transient cycles related to the SWOL fatigue crack growth 4.3 Prior to January analyses will be included in the existing plant transient monitoring program by 31,2011 January 31, 2011 to ensure that the actual plant transients do not exceed the SWOL fatigue analysis limits.

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