ML092100200
| ML092100200 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/11/2009 |
| From: | Singal B K Plant Licensing Branch IV |
| To: | Muench R A Wolf Creek |
| Singal, Balwant, 415-3016, NRR/DORL/LPL4 | |
| References | |
| TAC ME1393 | |
| Download: ML092100200 (7) | |
Text
UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 August 11, 2009 Mr. Rick A. Muench President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839 SUB..IECT: WOLF CREEK GENERATING STATION -REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PERMANENT ALTERNATE REPAIR CRITERIA LICENSE AMENDMENT REQUEST (TAC NO. ME1393)
Dear Mr. Muench:
By letter dated June 2, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML091590170), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) submitted a license amendment request to revise the Wolf Creek Generating Station Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10, "Steam Generator Tube Inspection Report." The licensee proposed to change the inspection scope, repair, and reporting requirements.
The proposed changes would establish permanent alternate repair criteria for portions of the SG tubes within the tubesheet.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information provided by the licensee and determined that additional information identified in the enclosure to this letter is needed in order for the NRC staff to complete the review. The draft copy of the request for additional information was provided to Mr. Steve Wideman of your staff via e-mail on July 23, 2009. WCNOC did not request further discussion to clarify the request for additional information and agreed to provide the response within 30 days of the date of the letter. Sincerely, "'-, \--i 1c6, '0 Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure As stated cc w/encl: Distribution via Listserv REQUEST FOR ADDITIONAL INFORMATION REGARDING PERMANENT H* ALTERNATE REPAIR CRITERIA FOR STEAM GENERATOR INSPECTIONS WOLF CREEK GENERATING STATION DOCKET NO. 50-482 By letter dated June 2, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML091590170), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee), submitted a license amendment request (LAR) to revise the technical specifications (TS) of Wolf Creek Generating Station (WCGS). The LAR proposed changes to the inspection scope and repair requirements of TS 5.5.9, "Steam Generator (SG) Program," and reporting requirements of TS 5.6.10, "Steam Generator Tube Inspection Report." The proposed changes would establish permanent alternate repair criteria for portions of the SG tubes within the tubesheet.
The U.S. Nuclear Regulatory Commission (NRC) staff has determined that additional information is needed in order to complete its review. The staff also notes that its review of Reference 1 is still ongoing, and NRC staff may have additional questions.
The Westinghouse Electric Company LLC (Westinghouse) document, WCAP-17071-P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," (Reference 1), was submitted with the June 2, 2009, letter, in support of the LAR. Reference 1! page 6-21! Table 6-6: This table contains a number of undefined parameters and some apparent inconsistencies with Table 5-2 on page 5-6. Please define the input parameters in Table 6-6. Reference 1, Section 6.2.2.2: Please explain why the finite element analysis was not run directly with the modified temperature distribution rather than running with the linear distribution and scaling the results. Reference 1, Section 6.2.3: Please explain why radial displacement is the "figure of merit" for determining the bounding segment. Does circumferential displacement not enter into this? Why is the change in tUbe hole diameter not the "figure of merit"? Reference 1! page 6-69: In Section 6.2.5.3, it is concluded that the tube outside diameter and the tubesheet tube bore inside diameter always maintain contact in the predicted range of tubesheet displacements.
However, for tubes with through-wall cracks at the H* distance, there may be little or no net pressure acting on the tube for some distance above H*. In Tables 6-18 and 6-19, the fourth increment in the step that occurs two steps prior to the last step suggests that there may be no contact between the tube and tubesheet, over a portion of the circumference, for a distance above H*. Is the conclusion in Section 6.2.5.3 valid for the entire H* distance, given the possibility that Enclosure
-2 the tubes may contain through-wall cracks at that location?
Additionally, please address the following issues: Clarify the nature of the finite element model ("slice" model versus axisymmetric SG assembly model) used to generate the specific information in Tables 6-1,2, and 3 (and accompanying graph entitled "Elliptical Hole Factors")
of Reference 6-15. What loads were applied? How was the eccentricity produced in the model? (By modeling the eccentricity as part of the geometry?
By applying an axisymmetric pressure the inside of the bore?) Explain why this model is not scalable to lower temperatures. Provide a table showing the maximum eccentricities (maximum diameter minus minimum diameter) from the 3 dimensional (3-D) finite element analysis for normal operating and steam line break (SLB). In Figure 2 of the White Paper, add plot for original relationship between reductions in contact pressure and eccentricity as given in Reference 6-15 in the graph accompanying Table 6-3. Explain why this original relationship remains conservative in light of the new relationship.
Explain the reasons for the differences between the curves. When establishing whether contact pressure increases when going from normal operating to SLB conditions, how can a valid and conservative comparison be made if the normal operating case is based on the original delta contact pressure versus eccentricity curve and the SLB case is based on the new curve? Reference 1, Section 6.3: Please verify if the previously calculated scale factors and delta D factors in Section 6.3 are conservative for (1) a steam line break (SLB) and a feedwater line break (FLB); (2) an intact divider plate assumption; and (3) all values of primary pressure minus crevice pressure that may exist along the H* distance for intact tubes and tubes with throughwall cracks at the H* distance. Reference 1, page 6-87: Please provide information on how the tube temperature (TT) on page 6-87 was determined.
For normal operating conditions, how is the TT assumed to vary as function of elevation? Reference 1, page 6-97, Figure 6-75: Contact pressures for nuclear plants with Model F SGs are plotted in Figure 6-75, but it is not clear what operating conditions are represented in the plotted data. Please clarify. Reference 1, page 6-112, Reference 6-5: This reference seems to be incomplete.
Please provide a complete reference. Reference 1, paoe 6-113, Reference 6-15: Table 6-3 in Reference 6-15 (SM-94-58, Revision 1) appears inconsistent with Table 6-2 in the same reference.
Please explain how the analysis progresses from Table 6-2 to Table 6-3.
-3Reference
- 1. oaae 8-9. Figure 8-1: There is an apparent discontinuity in the plotted data of the adjustment to H* for distributed crevice pressure.
Please provide any insight you may have as to why this apparent discontinuity exists. Reference
- 1. page 8-6, Section 8.1.4: Please clarify whether the "biased" H* distributions for each of the four input variables are sampled from both sides of the mean H* value during the Monte Carlo process, or only on the side of the mean H* value yielding an increased value of H*. Reference
- 1. page 8-14. Figure 8-6: The legend for one of the interactions shown between the coefficient of thermal expansion of the tube (aTS) and tubesheet (E TS) appears to contain a typographical error. Please review and verify that all values shown in the legend are correct. Reference 1, page 8-20. Case S-4: Why does the assumption of a 2-sigma value for the coe'fficient of thermal expansion of the tube (aT) and the tubesheet (aTS) to determine a "very conservative biased mean value of H*" conservatively bound the interaction effects between aT and a TS? Please describe how the "very conservative biased mean value of H*," as shown in Table 8-4, was determined. Reference
- 1. page 8-22. Case M-5: The description for this case seems to correspond to a single tube H* estimate rather than a whole bundle H* estimate.
Please explain how the analysis is performed for a whole bundle H* estimate. Reference
- 1. page 8-22: Case M-5 states, "Interaction effects are included because the 4.285 sigma variations were used that already include the effective interactions among the variables." Case M-5 also states that the 4.285 sigma variations come from Table 8-2; however, Table 8-2 does not appear to include interactions among the variables.
Please explain how the 4.285 sigma variations include the effect of interactions among the variables. Reference 1, page 8-22. Case M-6, first bullet: Please verify if the words "divided by 4.285" should appear at the end of the sentence. Reference 1, page 8-23, Case M-7: Please verify if the "2 sigma variation of all variables" was divided by a factor of 2. Reference 1, page 8-23, Case M-7: Please explain how this case includes the interaction effects between the two principal variables, aT and a TS. Reference
- 1. page 8-25, Table 8-4: Please explain why the mean H* calculated in the fifth case does not require the same adjustments, as noted by the footnotes, that all other cases in the table require. Reference
- 1. page 8-25, Table 8-4: Please verify the mean H* shown in the last case in the table.
-Section 8 of Reference 1: The variability of H* with all relevant parameters is shown in Figure 8-3. The interaction between aT and aTS are shown in Figure 8-5. Please explain why the direct relationships shown in these two figures were not sampled directly in the Monte Carlo analysis, instead of the sampling method that was chosen. Also, please explain why the sampling method chosen led to a more conservative analysis than directly sampling the relationships in Figures 8-3 and 8-5. As part of response, include discussion of main steam line break and whether it continues to be less limiting, from maximum H* perspective, than three times normal operating pressure. In the June 2, 2009, letter, WCNOC commits to monitor for tube slippage as part of the SG tube inspection program. The "due date/event" is prior to the start of refueling outage 1 R17. It is not clear whether the planned monitoring will be performed only once. Please modify the commitment to indicate that the tube slippage will be monitored during every SG tube inspection outage. In the June 2, 2009, letter, WCNOC commits to determine the position of the bottom of the expansion transition in relation to the top of the tubesheet and to enter "any significant deviation" into its corrective action program. This is a one-time verification prior to implementation of H*. Please modify the commitment to also include a commitment to notify the NRC staff if significant deviations in the location of the bottom of the expansion transition relative to the top of the tubesheet are detected. Reference 1, page 9-6, Section 9.2.3.1: The FLB heat-up transient is part of the plant design and licensing basis. Thus, it is the NRC staffs position that H* and the "leakage factors," as discussed in Section 9.4, should include consideration of this transient.
Please explain why the proposed H* and leakage factor values are conservative, even with consideration of the FLB heat-up transient.
As part of the response, address the FLB heatup transient in the Updated Safety Analysis Report, as this design-basis accident is part of the licensing basis. Please provide a rationale to justify basing the leakage factor on SLB, or commit to a leakage factor based on the feed line break heatup transient. During the staff review of the amendment request, it was noticed that the regulatory commitment regarding use of the leakage factor (see below) had been stated in the body of the document (page 10 of Attachment
- 1) but had been left off the list of regulatory commitments in Attachment V. Since the final leakage factor may change based on the FLB analysis (question 24 above), the proper factor will need to be used in the regulatory commitment.
For the condition monitoring (CM) assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.03 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment (DA), the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.03 and
-5 compared to the observed operational leakage. An administrative limit will be established to not exceed the calculated value.
References:
WCAP-17071-P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," dated April 2009 (ADAMS Accession Nos. ML091590169 and ML091590171, Proprietary Information.
Not Publicly Available).
August 11, 2009 Mr. Rick A. Muench President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839 WOLF CREEK GENERATING STATION -REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PERMANENT ALTERNATE REPAIR CRITERIA LICENSE AMENDMENT REQUEST (TAC NO. ME1393)
Dear Mr. Muench:
By letter dated June 2, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML091590170), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) submitted a license amendment request to revise the Wolf Creek Generating Station Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10, "Steam Generator Tube Inspection Report." The licensee proposed to change the inspection scope, repair, and reporting requirements. The proposed changes would establish permanent alternate repair criteria for portions of the SG tubes within the tubesheet.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information provided by the licensee and determined that additional information identified in the enclosure to this letter is needed in order for the NRC staff to complete the review. The draft copy of the request for additional information was provided to Mr. Steve Wideman of your staff via e-mail on JUly 23, 2009. WCNOC did not request further discussion to clarify the request for additional information and agreed to provide the response within 30 days of the date of the letter. Sincerely, /RAJ Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure As stated cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC LPLIV r/f RidsAcrsAcnw_MailCTR Resource RidsNrrDorlLpl4 Resource RidsNrrPMWolfCreek Resource RidsNrrLAJBurkhardt Resource RidsOgcRp Resource RidsRgn4MailCenter Resource AJohnson, NRR/DCIIICSGB RidsNrrDciCsgb Resource ADAMS Accession No'.. ML0921 00200 OFFICE NRR/LPL4/PM NRR/LPL4/LA DCI/CSGB/BC OGC NRRlLPL4/BC NRR/LPL4/PM NAME BSingal JBurkhardt MGavrilas Not Required MMarkley CFLyon for BSingal DATE 7/31/09 8/5109 8/5/09 8/11/09 8/11/09 OFFICIAL RECORD COPY