ML093520549

RAI Related to License Amendment Request to Update the Leak-Before-Break Evaluation for the Reactor Coolant Pump Section

ML093520549
Person / Time
Site:	Davis Besse
Issue date:	01/05/2010
From:	Sands S Plant Licensing Branch III
To:	Allen B FirstEnergy Nuclear Operating Co
Sands S,NRR/DORL, 415-3154
References
TAC ME2310
Download: ML093520549 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 5, 2010 Mr. Barry S. Allen Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST TO UPDATE THE LEAK-BEFORE-BREAK EVALUATION FOR THE REACTOR COOLANT PUMP SECTION AND DISCHARGE NOZZLE DISSIMILAR METAL WELDS (TAC NO. ME2310)

Dear Mr. Allen:

By letter to the Nuclear Regulatory Commission (NRC) dated September 28, 2009 (Agencywide Documents Access and Management System Accession No. ML092790438), FirstEnergy Nuclear Operating Company submitted a request to update the leak-before-break evaluation for the reactor coolant pump suction and discharge nozzle dissimilar metal welds.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on December 08, 2009, it was agreed that you would provide a response within 30 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3154.

Stephen P. Sands, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION DAVIS BESSE NUCLEAR POWER STAriON FIRSTENERGY NUCLEAR OPERATING COMPANY DOCKET NUMBER NO. 50-346 By letter dated September 28, 2009 (Agencywide Documents Access and Management System Accession No. ML092790438), FirstEnergy Nuclear Operating Company (the licensee),

submitted for Nuclear Regulatory Commission (NRC) review and approval a license amendment request to update the leak-before-break (LBB) evaluation as part of upcoming weld overlay installation on reactor coolant pump discharge and suction nozzles at Davis-Besse Nuclear Power Station during the spring 2010, refueling outage.

To complete its review of the Davis-Besse LBB evaluation, the staff requests the following information.

Enclosure B, Leak-before-Break Evaluation of Reactor Coolant Pump Suction and Discharge Nozzle Weld Overlays for Davis-Besse Nuclear Power Station,

1. Identify which weld overlay, the full structural weld overlay (FSWOL) or optimized weld overlay (OWOL), was used in the stress distributions as shown in Figures 3-3, 3-4, 3-5, and 3-6 (for the reactor coolant pump (RCP) inlet nozzle) and Figures 3-7,3-8,3-9,3-10 (for the RCP outlet nozzle).

2. The last paragraph on Page 3-2 states that an OWOL may still be applied to a dissimilar metal weld (DMW) which contains a flaw that has a depth of greater than 50 percent but less than 75 percent through-wall. The staff does not agree with applying an OWOL on a DMW with a pre-existing flaw that is greater than 50 percent through-wall. Discuss the basis for applying an OWOL on such a degraded DMW.

3. The third paragraph on Page 3-3 discusses thermal boundary conditions (wet or dry) during weld overlay installation. Water backing effects (cooling during the weld overlay laydown) and weld overlay sequencing may affect the weld residual stress model results. Discuss whether water cooling and weld sequencing effects were analyzed, or if not, what was the justification for not considering the effects of these conditions on the modeled residual stresses? Discuss whether weld overlay installation at the RCP inlet and outlet nozzles will be performed when the inside of pipe will be dry or with water.

4. The third paragraph on Page 3-8 states that "... As an alternative to the above requirements, for cases in which if current examination requirements are satisfied by inspecting the inner 1/3 of the original DMW from the inside diameter (10) of the nozzle, the utility may continue to perform such examinations, in lieu of the WOL examinations specified above. In such cases, the outside diameter (00) examination requirement is just the overlay itself, and is required only for the pre service inspection performed after WOL application ... " (a) It appears that if ultrasonic testing (UT) is performed from the 10 and 00 surfaces, the 00 inspection will only inspect the weld overlay and not the base metal. The staff believes that the outer 50 percent of the base metal thickness should also be inspected by UT from the 00 surface because the overall integrity of the pipe relies on both the OWOL and the outer 25 percent wall thickness of the base metal.

-2 The staff does not believe this requirement is included in the submitted weld overlay relief request. Clarify the above statement.

5. The first paragraph on Page 3-10 states that for a DMW containing a pre-existing flaw, the overlaid DMW shall be inspected once in the next 5 years. This is contrary to previously NRC approved relief requests which require that the overlaid DMW be inspected during either the first or second refueling outage after overlay installation. Assuming an operating cycle of 18 months, the overlaid DMW should be examined within 3 calendar years after weld overlay installation.

Justify the 5-year inspection interval.

6. Figure 3-11 showed the stress intensity factor vs. flaw depth for the RCP inlet after FSWOl installation. Figure 3-12 shows the stress intensity factor vs. flaw depth for the RCP outlet after OWOl installation. It appears that the OWOl installation provides more favorable stress intensity factor results than the FSWOL. Explain why the OWOl in Figure 3-12 shows more favorable stress intensity factor results than the FSWOl in Figure 3-11.

7. Table 3-1 presented the length of time for a postulated initial flaw size to grow to the design flaw size after weld overlay installation. For the RCP outlet nozzle, under the OWOl application, the postulated initial axial flaw of 50 percent through-wall of the pipe thickness will take more than 60 years to grow to the design flaw size. Based on the Davis-Besse relief request, the initial axial flaw should be 75 percent through-wall because the ultrasonic examination has not been qualified to examine outer 50 percent wall thickness of the base metal. The final flaw size for this case should be 100 percent through-wall. (a) Clarify why the initial axial flaw size for RCP outlet nozzle under the OWOl design is assumed to be 50-percent through-wall. (b)

Confirm that the initial circumferential flaw under the OWOl design is 50 percent through-wall and the final flaw size is 75 percent through-wall. (c) Based on footnote NO.3 of Table 3-1, the initial circumferential flaw is assumed to be 75 percent through-wall. However, in the Davis Besse relief request, the initial circumferential flaw is assumed to be 50-percent through-wall.

Clarify whether there is a discrepancy in the initial circumferential flaw size between the lBB license amendment and the overlay relief request.

8. The submittal used an axi-symmetric model to analyze residual stresses in the DMW and has been shown to produce conservative results at the ID, but not necessarily at the OD. The RCP discharge and suction nozzles are connected to an elbow. The elbow is a stress concentration location because of its configuration and the weld overlay will be installed on a portion of the elbow. Figures 3-1 and 3-2 included the elbow in the finite element model. (a) Discuss the impact of the weld overlay on the residual stresses in the elbow. (b) Discuss any precautions taken in your welding procedures for the weld overlay installation on the elbow to minimize the potential for over-stress of the elbow.

9. Figure 4-1 showed the crack paths assumed at four locations for the inlet and outlet nozzles.

(a) Discuss why a crack path was not assumed at the middle of the DMW in Figure 4-1.

(b) Confirm that for the crack in each path, the material properties of each individual component (Le., cast austenitic stainless steel RCP nozzles, stainless steel safe end, Alloy 82/182 weld, and ferrite elbow) were used to calculate critical crack size and leak rates for each specific crack path.

3

10. Section 4.2 stated that the RCP nozzle loads used in the current lBB evaluation were taken from the AREVA Engineering Information Record, 51-9094884-000, dated October 21,2008.

However, it appears that the RCP nozzle loads in the AREVA report do not include the impact of the weight of weld overlay (Le., forces and moments generated by the weight of the weld overlay). If the weight of weld overlay is not included in the RCP nozzle loads in the current lBB evaluation, discuss the validity of the critical crack size, leakage crack size, and associated safety margins.

11. Section 4.3.3, page 4-3, discussed thermal embritllement of cast austenitic stainless steel (CASS) material in the RCP nozzles. (a) Discuss whether the saturated fracture toughness of the CASS was used in the current lBB evaluation. (b) Provide the saturated fracture toughness of the CASS nozzles used in the analysis.

12. Table 4-2 presented the pipe loads at the DMW. Axial shrinkage of the overlay can cause a tensile axial stress in the rest of the system when the weld overlay is in situ with the pipe system connected to the vessel and steam generator. This shrinkage should result in slightly different thermal stresses at the DMW than the original piping stress analysis. Discuss whether the shrinkage stresses were accounted for in the flaw stability calculation and in the leakage calculation.

13. Clarify whether residual stresses calculated in Section 3 of the report were used or were involved in the flaw stability and leakage calculations?

14. Section 4.3.2 discussed the J-T curve for ferritic materials. Section 4.3.3 discussed the lower bound J-T curve for CASS material. (a) Explain which J-T curve (ferritic or CASS) was used in the flaw stability calculation. (b) Discuss why the J-T curve for Alloy 52M overlay material or Alloy 82/182 DMW was not mentioned in Section 4.0 and it appears that they are not used in the flaw stability calculation.

15. The first paragraph on Page 5-3 stated that the larger Z-factor for the carbon steel material from the American Society of Mechanical Engineers (ASME) Code will be conservatively used in the critical flaw evaluation. Table 5-1 showed that the Z-factor for the carbon steel elbow is 1.82.

Discuss whether 1.82 was used in the critical flaw evaluation for all the materials (Le., nozzle, safe end, DMW, overlay, and elbow) in crack path 1,2,3, and 4 as shown in Table 5-2, because the Z-factor of 1.82 is not applicable to nozzle, safe-end and DMW, which are not made of carbon steel.

16. Table 5-2 showed the critical crack size. (a) Confirm that the through-wall critical crack length in the circumference direction is assumed to be the same in the DMW as in the weld overlay. (b) Confirm that for the leak rate calculation, the crack size in the base metal and in the overlay is assumed to be the same (page B-7). (c) Discuss why there is not much difference in the critical crack size for the FSWOl and the OWOl design at RCP discharge nozzle.

17. Page 2-2 shows the half critical crack sizes calculated for the base metal and weld metal in the original lBB evaluation. The original lBB evaluation showed that the critical crack sizes for the base metal and weld metal (without the weld overlay) are 20.36 inches (10.18" x 2) and 37.08 inches (18.54" x 2), respectively. The differences between these two critical crack sizes

4 are substantial (20.36" vs. 37.08"). (a) Explain why the critical crack sizes calculated for the weld overlay as shown in Table 5-2 in the current LBB evaluation do not show the same large differences as in the original LBB calculation. (b) The increase in critical crack size from the original LBB evaluation (20.36" and 37.08") to the critical crack sizes after weld overlay installation as shown in Table 5-2 is substantial. For example, for the base metal case, the critical crack size increase is about 60 percent (from 20.36" to 50.24"). For the weld metal case, the critical crack size increase is about 35 percent (from 37.08" to 57.89"). It is expected that the critical crack size will be increased after the weld overlay installation. However, it appears that the percentage of critical crack size increase exceeded the percentage of wall thickness increase between the original pipe and the overlaid pipe. Discuss the contributors to the increase in critical crack size after the weld overlay installation.

18. Table 6-3 presented leak rates for a leakage flaw size equal to half of the critical flaw size.

Explain why the leak rates for cracks in Paths 1 and 4 in Table 6-3 are much higher (as much as 7 times) than the leak rates in Paths 2 and 3, even though the critical crack sizes in Paths 1, 2, 3, and 4 are all about the same.

19. Table 6-5 presented a comparison of the leakage flaw size for the 10 gallons per minute (gpm) leak rate between the original Babcock & Wilcox LBB evaluation and the current LBB evaluation. However, the staff cannot find the same leakage flaw size under the current evaluation column in Table 6-5 among the leakage flaw sizes in Table 6-2 (which shows the leakage flaw size for 10 gpm leak rate) of the submittal. Explain where and how the leak rates in Table 6-5 were calculated or taken from.

20. The first sentence on Page A-3 stated that, optionally, the effect of internal pressure on the crack surface of both base material and weld overlay can be evaluated. Discuss whether internal pressure was applied to the crack surface of both base material and overlay in calculating the flaw stability.

21. Section B.3 discusses the effect of crack face pressure on leakage. Clarify whether crack face pressure was considered in the final leak rate results.

22. Item 4.d of Section 6.2 states that: "Crack roughness is taken as 0.000197 inches for fatigue cracking in materials other than the Alloy 82/182 weld. There are no turning losses assumed for fatigue cracking." Section 6.2 did not mention roughness for primary water stress corrosion cracking (PWSCC) in the DMW. For the DMW, Item 5 of Section 6.2 indicates that the crack morphology properties for the PWSCC-susceptible Alloy 82/182 material were taken from Appendix B of the LBB evaluation. Item 5 of Section 6.2 states that "For the weld with Alloy 82/182 material, the adverse effects of PWSCC crack morphology will be considered for the affected material as described in Appendix B; for other material, the crack morphology for fatigue cracking is used ...". NRC staff believes that the fatigue crack morphology properties quoted in Item 4.d of Section 6.2 are significantly lower than the values reported in NUREG/CR-6004 and numerous prior Westinghouse LBB submittals. The lower roughness and number of turns used in the analysis will result in a shorter postulated leakage crack size, hence, the margin between critical crack length and leakage crack length would be overstated with respect to an analysis where one used the NUREG/CR-6004 values. (a) Justify the use of the crack morphology parameters, e.g. roughness values and number of turns, used in the leak rate calculation.

5 (b) Discuss how the PWSCC and fatigue properties were combined to yield a single set of composite crack morphology parameters. (c) Provide the roughness for PWSCC. (d) Provide the average number of turns for a typical PWSCC crack. (e) Discuss how many 45 and gO-degree turns were modeled in a typical PWSCC crack.

23. Discuss whether there are laboratory experiments which have been performed to verify the accuracy of the analytical method that the licensee used (Le., PICEP) to predict the leak rates from through-wall crack in the overlaid DMW. If no experiments were performed, justify the accuracy of the leak rate methodology and results.

24. The leaking coolant flows from the inside surface of the pipe through the postulated PWSCC crack in the DMW and the postulated fatigue crack in the weld overlay to the outside surface of the pipe. It is not clear how the final leak rate was calculated based on the leak rate in the PWSCC crack in the DMW and the leak rate in the fatigue crack in the weld overlay. The leak rate in the crack in the DMW should be slower than the leak rate in the fatigue crack in the base metal. (a) Discuss how the leak rate in the DMW crack is combined with the leak rate in the weld overlay fatigue crack to derive a final leak rate. (b) Discuss whether the crack opening displacement (COD) in the PWSCC crack is the same as in the fatigue crack. If the COD are not the same, discuss how the final COD is calculated.

25. Section 8.3.1, Item 1 discusses fatigue cracks in a baseline PICEP run. (a) Discuss whether PWSCC cracks in the DMW are also analyzed in the baseline PICEP run. (b) Discuss why the baseline run produces 20 leakage calculations for increasing crack size.

26. Section 8.3.1, Item 2 discusses the second set of computer runs which produces modified crack opening displacement as a function of crack size. (a) The staff presumes that as crack size increases, crack opening displacement increases which in turn increases leak rate. Is this the correct observation? (b) Item 2 states that the leakage flow rate for the original crack parameters for the increased crack opening is generated. Clarify what are "original crack parameters."

27. Under normal operation conditions of the RCP piping, the leakage may occur in a two-phase condition (Le., a mixture of steam and water) at the exit. Clarify whether the two-phase flow condition has been considered in the leak rate calculation as such. If a two phase flow is assumed, discuss how the leak rate of steam is converted to gpm.

January 5, 2010 Mr. Barry S. Allen Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST TO UPDATE THE LEAK-BEFORE-BREAK EVALUATION FOR THE REACTOR COOLANT PUMP SECTION AND DISCHARGE NOZZLE DISSIMILAR METAL WELDS (TAC NO. ME2310)

Dear Mr. Allen:

By letter to the Nuclear Regulatory Commission (NRC) dated September 28,2009 (Agencywide Documents Access and Management System Accession No. ML092790438), FirstEnergy Nuclear Operating Company submitted a request to update the leak-before-break evaluation for the reactor coolant pump suction and discharge nozzle dissimilar metal welds.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on December 08, 2009, it was agreed that you would provide a response within 30 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3154.

Sincerely, IRA by M. Mahoney fori Stephen P. Sands, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv DISTRIBUTION:

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