ML101170016
| ML101170016 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 04/23/2010 |
| From: | Becker J Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-10-045, TAC ME2995, TAC ME2996 | |
| Download: ML101170016 (9) | |
Text
{{#Wiki_filter:Pacific Gas and Electric Company April 23, 2010 PG&E Letter DCL-1 0-045 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 James R. Becker Site Vice President Diablo Canyon Power Plant Mail Code 104/5/601
- p. O. Box 56 Avila Beach, CA 93424 805.545.3462 Internal: 691.3462 Fax: 805.545.6445 10 CFR 50.90 Response to NRC Request for Additional Information Regarding License Amendment Request 09-06, "Critical Damping Value for Structural Dynamic Qualification of the Control Rod Drive Mechanism Pressure Housings" (TAC Nos.
ME2995 and ME2996)
Reference:
- 1. PG&E Letter DCL-09-86, "License Amendment Request 09-06, "Critical Damping Value for Structural Dynamic Qualification of the Control Rod Drive Mechanism Pressure Housings," dated December 14, 2009. (ADAMS Accession No. ML093580092)
Dear Commissioners and Staff:
By letter dated December 14, 2009 (Reference 1), Pacific Gas and Electric Company (PG&E) submitted a license amendment request to revise the licensing basis and the Final Safety Analysis Report Update (FSARU) to allow use of a damping value of 5 percent of critical damping for the structural dynamic qualification of the control rod drive mechanism pressure housings on the replacement reactor vessel head for the design earthquake, double design earthquake, Hosgri earthquake, and loss-of-coolant accident (LOCA) loading conditions. On April 15, 2010 (ADAMS Accession No. ML101050521), the NRC staff requested additional information required to complete the review of LAR 09-06. PG&E's responses to the staff's questions are provided in the enclosure. This information does not affect the results of the technical evaluation or the no significant hazards consideration determination previously transmitted in Reference 1. PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter. This letter includes no revisions to existing regulatory commitments. A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak
- Diablo Canyon
- Palo Verde
- San Onofre. South Texas Project
- Wolf Creek
I~ Document Control Desk April 23, 2010 I & Page 2 PG&E Letter DCL-10-045 If you have any questions, or require additional information, please contact Tom Baldwin at (805) 545-4720. I state under penalty of perjury that the foregoing is true and correct. Executed on April 23, 2010. SinC~eIY, ~c-- James R. Becker Site Vice President tcg4231 50304279 Enclosure cc: Diablo Distribution cc/enc: Gary W. Butner, Acting Branch Chief, California Department of Public Health Elmo E. Collins, NRC Region IV Michael S. Peck, NRC, Senior Resident Inspector Alan B. Wang, Project Manager, Office of Nuclear Reactor Regulation A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak
- Diablo Canyon
- Palo Verde
- San Onofre. South Texas Project
- Wolf Creek
Enclosure PG&E Letter DCL-1 0-045 PG&E Response to NRC Request for Additional Information Regarding License Amendment Request (LAR) 09-06, "Critical Damping Value for Structural Dynamic Qualification of the Control Rod Drive Mechanism Pressure Housings" NRC Question 1: Please provide the following information: (a) Provide a summary of the recent analysis results for the new CRDM pressure housing based on a 5% critical damping value for all load combinations that contain DE, DOE, HE, and LOCA loadings, showing the respective margins. (b) Provide a brief description of the structural dynamic analysis methodology used in the qualification of the CRDM pressure housings on the replacement reactor vessel head. (c) Provide, for comparison, a summary of lower damping values used in the original CRDM analysis and a summary of the design basis analysis results, with the respective margins. (d) Clarify that the higher 5% critical damping value will only be applied to the CRDMs, and will not be applied to the reactor coolant piping or reactor internals analyses. PG&E Response: (a) The following tables summarize the recent analysis results for the new control rod drive mechanism (CRDM) pressure housing based on a 5 percent critical damping value for all load combinations that contain design earthquake (DE), double design earthquake (DOE), Hosgri earthquake (HE), and loss-of-coolant accident (LOCA) loadings. The interaction ratio is the calculated value divided by the allowable value. Interaction ratios below 1.0 represent margin. The values are bounding for both units. Table 1: Rod Travel Housing Stresses Calculated Allowable Interaction ASME Classification Value Value Ratio Local Primary Membrane + Primary Bending 13.06 ksi 24.3 ksi 0.54 Stress (DE Load Combination) Local Primary Membrane + Primary Bending + 39.74 ksi 48.6 ksi 0.82 Secondary Stress (DE Load Combination) Fatigue Usage Factor (DE Load Combination) .004 1.0 0.004 Local Primary Membrane + Primary Bending 36.91 ksi 58.3 ksi 0.63 Stress (DOE + LOCA Load Combination) Local Primary Membrane + Primary Bending 17.32 ksi 58.3 ksi 0.30 Stress (DOE Load Combination) 1
Calculated ASME Classification Value Local Primary Membrane + Primary Bending 28.77 ksi Stress (HE Load Combination) Table 2: Latch Housing Stresses Calculated ASME Classification Value Local Primary Membrane + Primary Bending 22.01 ksi Stress (DE Load Combination) Local Primary Membrane + Primary Bending + 18.71 ksi Secondary Stress (DE Load Combination) Note: Secondary Stress Excludes Thermal Bending Stress per ASME Simplified Elastic-Plastic Analysis Rules Fatigue Usage Factor (DE Load Combination) 0.033 Local Primary Membrane + Primary Bending 54.74 ksi Stress (DOE + LOCA Load Combination) Local Primary Membrane + Primary Bending 39.29 ksi Stress (DOE Load Combination) Local Primary Membrane + Primary Bending 49.23 ksi Stress (HE Load Combination) Enclosure PG&E Letter DCL-1 0-045 Allowable Interaction Value Ratio 58.3 ksi 0.49 Allowable Interaction Value Ratio 24.3 ksi 0.91 48.6 ksi 0.38 1.0 0.033 58.3ksi 0.94 58.3 ksi 0.67 58.3 ksi 0.84 Table 3: CRDM Nozzle between Top of Head and Dissimilar Metal Weld Calculated Allowable Interaction ASME Classification Value Value Ratio Local Primary Membrane + Primary Bending 26.64 ksi 34.9 ksi 0.76 Stress (DE Load Combination) Local Primary Membrane + Primary Bending + 66.77 ksi 69.9 ksi 0.96 Secondary Stress (DE Load Combination) Fatigue Usage Factor (DE Load Combination) 0.003 1.0 0.003 Local Primary Membrane + Primary Bending 71.43 ksi 83.8 ksi 0.85 Stress (DOE + LOCA Load Combination) Table 4: CRDM Nozzle at J-Groove Weld Calculated Allowable Interaction ASME Classification. Value Value Ratio Local Primary Membrane + Primary Bending 25.88 ksi 35.0 ksi 0.74 Stress (DE Load Combination) Local Primary Membrane + Primary Bending + 66.77 ksi 69.9 ksi 0.96 Secondary Stress (DE Load Combination) Fatigue Usage Factor (DE Load Combination) 0.282 1.0 0.282 Local Primary Membrane + Primary Bending 67.47 ksi 83.9 ksi 0.80 Stress (DOE + LOCA Load Combination) 2
Enclosure PG&E Letter DCL-1 0-045 (b) The following provides a brief description of the structural dynamic analysis methodology used in the qualification of the CRDM pressure housings on the replacement reactor vessel head: The DE and DOE horizontal direction seismic loads, acting on the CRDMs, were determined using finite element models and nonlinear seismic time history methods. For Unit 2, a center row of five CRDMs was modeled to include the effects of different CRDM lengths due to the curvature of the reactor head. For Unit 1, a center row of seven CRDMs was modeled. This approach is conceptually consistent with the approach used to determine seismic loads in the original CRDM seismic analysis. The finite element models included nonlinear spring elements to represent the gaps and contact between the seismic plates (near the tops of the CRDM rod travel housings) and to represent the contact between the outer seismic plates and the seismic stop plates. A linear spring was also included to represent the combined tie-rod and Integrated Head Assembly (lHA) stiffness connection to the wall. The IHA was included in the models as a vertical beam connecting the Reactor Vessel Closure Head (RVCH) to the tie-rod spring. An IHA horizontal beam was included to connect the IHA vertical beam to the seismic stop plates. Time history input motions for the CRDM DE and DOE horizontal earthquake analyses were developed at the RVCH elevation and at the elevation corresponding to the tie-rod/IHA connection to the wall. The time history input motions at the tie-rod/IHA connection elevation were used to generate acceleration response spectra for comparison to existing response spectra at this same elevation. The generated response spectra enveloped the existing response spectra at all frequencies within the CRDM response frequency range. Time history input motions at the RVCH were developed by Westinghouse for use by AREVA who supplied and qualified the replacement RVCH, IHA, and CRDMs. The time history inputs and CRDM models were used with the ANSYS structural analysis program to determine seismic forces acting on the CRDMs along their lengths due to DE and DOE horizontal direction earthquakes. The analyses used ANSYS Beta damping corresponding to 5 percent of the critical damping ratio for the CRDMs. Analyses show the CRDM fundamental frequency range is 5 to 8 Hz and the dominant frequencies of response are below 10Hz. Therefore, a conservative value of 10Hz was used to determine the CRDM Beta damping value used in the analyses. The HE horizontal direction seismic forces and the vertical direction seismic forces for DE, DOE, and HE were determined using linear elastic response spectra methods. The response spectra analyses included the center CRDM (the shortest) and a peripheral CRDM (the longest) to envelope the results for all CRDMs. Vertical direction 5 percent damping response spectra at the RVCH elevation were used for 3
Enclosure PG&E Letter DCL-1 0-045 the vertical direction seismic analyses since the CRDMs are supported vertically only at the RVCH elevation. The HE horizontal direction seismic analysis used 5 percent damping response spectra at the RVCH elevation and the elevation of the seismic plates. To account for gaps between the seismic plates, a static analysis was also performed using a forced horizontal direction displacement at the CRDM seismic plate elevation that corresponds to the maximum cumulative gap. The resulting seismic forces acting on the CRDMs due to both vertical and horizontal DE, DOE, and HE earthquakes were used with CRDM physical properties to determine seismic stresses acting on the CRDM. The seismic stresses were combined with other applicable stresses such as pressure and thermal to determine the total combined stress for each applicable load combination. The combined stresses were determined at each critical location along the length of the CRDM assembly including locations along the rod travel housing, latch housing, and CRDM penetration into the RVCH. These stresses were then evaluated for acceptance per Section NB-3200 of the ASME Boiler and Pressure Vessel Code 2001 Edition through 2003 Addendum. This Code was satisfactorily reconciled to the original CRDM code of record. (c) The original CRDM analysis was performed by Westinghouse using the nonlinear time history analysis method with stiffness damping corresponding to 3 percent of the critical damping. DOE loads were determined to be larger than the HE loads and therefore the DOE results bound the HE results. The following table summarizes results from the original CRDM analysis along with allowable values and interaction ratios. The interaction ratio is the calculated value divided by the allowable value. Interaction ratios below 1.0 represent margin. Table 5: Rod Travel Housing and Latch Housing Loads Calculated Allowable Interaction Location and Load Value Value Ratio Rod Travel Housing DDE Moment 64 in-kip 73 in-kip 0.88 CRDM Latch Housing DDE Moment 140 in-kip 670 in-kip 0.21 Table 6: CRDM Nozzle between Top of Head and Dissimilar Metal Weld Calculated Allowable Interaction ASME Classification Value Value Ratio Primary Membrane + Primary Bending Stress 23.67 ksi 23.98 ksi 0.99 (DE Load Combination) Primary Membrane + Primary Bending + 68.40 ksi 69.9 ksi 0.98 Secondary Stress (DE Load Combination) Fatigue Usage Factor (DE Load Combination) 0.002 1.0 0.002 Local Primary Membrane + Primary Bending 55.73 ksi 58.32 ksi 0.96 Stress (DDE + LOCA Load Combination) 4
Enclosure PG&E Letter DCL-10-045 Table 7: CRDM Nozzle at J-Groove Weld Calculated Allowable Interaction ASME Classification Value Value Ratio Local Primary Membrane + Primary Bending 24.35 ksi 34.48 ksi 0.71 Stress (DE Load Combination) Local Primary Membrane + Primary Bending + 50.21 ksi 69.9 ksi 0.72 Secondary Stress (DE Load Combination) Fatigue Usage Factor (DE Load Combination) 0.191 1.0 0.191 Local Primary Membrane + Primary Bending 56.34 ksi 83.88 ksi 0.67 Stress (DOE + LOCA Load Combination) (d) The higher 5 percent critical damping value per this LAR will only be applied to the CRDMs and will NOT be applied to the reactor coolant piping or reactor internals analyses. The original analysis seismic model included the reactor vessel and internals along with the CRDMs and seismic support platform. As discussed above, the new seismic model includes the CRDMs and the IHA that incorporates a seismic support platform but the model ends at the reactor vessel closure head. Time History and response spectra inputs are developed by Westinghouse at the reactor vessel closure head for use in the new CRDM analysis. Regulatory Guide 1.61 Revision 1 damping is used for the IHA and associated components. As discussed with the NRC staff, a LAR will be submitted to obtain NRC approval to use these damping values. Damping values for the reactor vessel internals and reactor piping are unchanged from the original analyses. NRC Question 2: Pages 3 and 4 along with Figure 1 of the enclosure to PG&E's letter dated December 14, 2010 (DCL-09-086), provide the basis for similarity between the original equipment manufacturer (OEM) CRDM and the new AREVA supplied replacement CRDM from considerations such as the seismic support plate gap, CRDM length, latch housing and rod travel housing materials, CRDM geometrical section properties, operating parameters, appurtenances, and design differences. Provide a similar comparison of the AREVA-supplied replacement CRDMs and the CRDMs tested by Westinghouse, as described in the topical report, WCAP-7921-AR. PG&E Response: The following provides the cQmparison of the AREVA-supplied replacement CRDMs and the CRDMs tested by Westinghouse, as described in the topical report, WCAP-7921-AR:
- 1. Gap at the seismic support plates. The AREVA-supplied replacement CRDMs provide a 0.070 inch gap (cold) and a 0.100 inch gap (hot) between the CRDM seismic support plates. The Westinghouse CRDM testing determined the damping 5
Enclosure PG&E Letter DCL-1 0-045 values for three different gap values: zero gap, 0.015 inch gap, and 0.060 inch gap. The testing showed larger gaps significantly increased the measured CRDM damping. The seismic support plate gaps for the AREVA-supplied CRDMs exceed the maximum gaps included in the Westinghouse testing to ensure the test results are applicable to the replacement CRDM installation.
- 2. CRDM length. The Westinghouse CRDM damping testing utilized a model L-1058 CRDM that is designed for a 10-foot core. Diablo Canyon has 12-foot cores and therefore the AREVA-supplied CRDMs are approximately 17 to 40 inches longer than the tested L-1058 CRDM as measured from the top of the reactor vessel closure head to the top of the rod travel housing. See Figure 1. It is expected that the longer latch assembly and rod travel housing of the AREVA-supplied CRDMs will have a positive effect on damping (ie: higher damping when compared to the L-1058 tested CRDM).
- 3. Latch Housing and Rod Travel Housing (RTH) Materials. The AREVA-supplied CRDMs and the L-1058 tested CRDM both use type 304 stainless steel.
- 4. Section Properties. Figure 1 shows section property differences between the L-1058 CRDMs that were tested and the AREVA-supplied CRDMs. The original Diablo Canyon L-106A CRDMs supplied by Westinghouse have essentially the same section properties as the AREVA-supplied CRDMs. Westinghouse has determined these section property differences will have negligible impact on CRDM damping.
Other U.S. plants base their current license basis acceptance to use 5 percent damping for their CRDM analysis on the Westinghouse L-1 058 testing.
- 5. Operating Parameters. The Westinghouse L-1058 CRDM damping tests were performed at a pressure of 2,250 psi and a temperature of 400 degrees Fahrenheit.
Normal operating conditions included in the AREVA-supplied CRDM design are 2,235 psig and 610 degrees Fahrenheit. The difference in temperature is expected to have negligible effect on damping.
- 6. Appurtenances. The AREVA-supplied CRDMs have the typical appurtenances including coil assemblies, digital rod position indicator coil stack, and an attached control rod. The Westinghouse L-1058 testing included an operating coil stack, a nonfunctioning rod position indicator coil stack and a weight at the bottom of the drive shaft to simulate the weight of the control rod.
- 7. Specific Design Differences. The specific design differences provided in DCL 086 between the AREVA-supplied CRDMs and the CRDMs originally installed at Diablo Canyon are also applicable between the AREVA-supplied CRDMs and the L-1058 CRDMs that were tested. These differences have negligible effect on damping. In addition, there are specific differences in CRDM length and section properties as discussed above.
6
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