ML14036A359
| ML14036A359 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 02/05/2014 |
| From: | Ameren Missouri, Union Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14036A358 | List: |
| References | |
| TAC ME7708, ULNRC-06079 | |
| Download: ML14036A359 (7) | |
Text
ULNRC-06079 February 5, 2014 Page 1 of 5 CALLAWAY PLANT UNIT 1 LICENSE RENEWAL APPLICATION SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
SET #29, RAI B2.1.6-4d, PART (a)
ULNRC-06079 February 5, 2014 Page 2 of 5 RAI B2.1.6-4d
Background:
Generic Background Information-The Nuclear Regulatory Commission's (NRC's) position regarding implementation of recommended inspection and evaluation (I&E) criteria from the MRP-227-A report as part of a plant-specific aging management program (AMP) for reactor vessel internal (RVI) components is given in NRC Regulatory Issue Summary (RIS) No. 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," dated July 21, 2011. The RIS recommends that the review of the I&E bases for Category D pressurized-water reactor (PWR) facilities be assessed as part of the review of the applicant's AMP for its RVI components, including the bases for resolving any applicant/licensee action items (A/LAIs) on the MRP-227-A I&E methodology that are applicable to the design of the RVI components at the facility. These A/LAIs are identified in the NRC's revised safety evaluation (SE, Rev. 1, dated December 16, 2011) on the MRP-227-A I&E methodology. According to RIS No. 2011-07, Callaway Plant, Unit 1 (Callaway) is categorized as a Category D facility, which applies to PWR applicants that either will be submitting a license renewal application (LRA) that is based on the recommended criteria in NUREG 1801, "Generic Aging Lessons Learned (GALL) Report," Revision 2, or currently have GALL Report Revision 2 based LRAs pending an NRC review.
Plant-Specific Background Information-The staff's understanding is that the current licensed core power level for Callaway is set at 3565 MWt, as approved in the NRC's license amendment and safety evaluation of March 30, 1988, which was issued on the 4.5 percent stretch power uprate request for Callaway (ADAMS Accession No. ML021650524).
In A/LAI No. 1, the staff requested that applicants with a PWR design provide a demonstration that the bases and assumptions for the I&E methodology in Topical Report MRP-227-A are applicable and bounding for the design of the RVI components at the applicant's plant. The applicant responded to the request in A/LAI No. 1 in the applicant's response to RAI B2.1.6-4a which was provided in Ameren Letter No. ULNRC-05950, dated January 24, 2013.
In its January 24, 2013, response letter to RAI B2.1.6-4a, the applicant provided the following LRA commitment (as given in Commitment No.4 in LRA UFSAR Supplement Table A4-1) as the basis for resolving the request in A/LAI No. 1:
Each applicant/licensee is responsible for assessing its plant's design and operating history and demonstrating that the approved version of MRP-227 is applicable to the facility. Each applicant/licensee shall refer, in particular, to the assumptions regarding plant design and operating history made in the FMECA and functionality analyses for reactors of their design (i.e., Westinghouse, CE, or B&W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227.
Issue:
Since Callaway is a RIS 2011-07 Category D plant, the resolution of A/LAI No. 1 needs to be resolved as part of the staff's review of the Callaway LRA and PWR Vessel Internals Program.
ULNRC-06079 February 5, 2014 Page 3 of 5 Request:
(a)
Clarify whether the design of RVI components at Callaway includes any non-welded or bolted austenitic stainless steel components whose design stresses are greater than 30 ksi and whose materials were cold worked to 20 percent or greater cold-work levels. If so, justify why the current I&E bases in MRP-227-A report are sufficient to provide for management of cracking or other applicable aging effects in these non-welded components. Otherwise, clarify and justify how the MRP-227-A report l&E bases for these RVI components will be adjusted as part of the applicant's response to the NRC's request in A/LAI No. 2.
(b)
Clarify whether Ameren Missouri has ever utilized an atypical fuel design or fuel management protocols that could make the assumptions in MRP-227-A on core design, core loading, and core leakage patterns non-representative for the Callaway RVI design, including those that might have been approved for the facility under the NRC's process for reviewing power uprate/power change license amendment requests. If so, justify why the current I&E bases in MRP-227-A report are sufficient to provide for management of cracking and other applicable aging effects in the plant's RVI components based on the actual fuel loading patterns and fuel power conditions that are approved in the current licensing basis. Otherwise, clarify and justify how the MRP-227-A report I&E bases for these RVI components will be adjusted as part of the applicant's response to the NRC's request in A/LAI No. 2.
Callaway Response (a) The response to RAI B2.1.6-4d part (a) will be submitted at a later date.
Callaway commits to perform one or more of the options identified by LRA Table A4-1 item
- 44. Callaway has conservatively included all applicable MRP-191 Table 4-4 components as potentially subject to 20% or greater cold work and 30 ksi operating stress. Callaway may revise this commitment at a future date if any of the applicable MRP-191 Table 4-4 components are determined not to be subject to 20% or greater cold work and 30 ksi operating stress. LRA Table A4-1 item 44 has been added as shown in Amendment 30 in Enclosure 2.
(b) MRP-227-A assumed that the degradation rate of the reactor internals would decrease during the second 30 years of operation. This requires the use of low leakage reactor cores during this period, and thus precludes the use of out-in core loading patterns. EPRI letter MRP 2013-025, Attachment 1, provides criteria for Combustion Engineering and Westinghouse PWRs with regard to radial boundary limitations, upper axial boundary limitations, and lower axial boundary limitations which, if met, ensure that the assumptions of MRP-227-A are met. As stated in MRP 2013-025, these criteria apply to operation going forward; i.e., during the second 30 years of operation.
To meet the radial boundary limitations, the following limits must be met:
Heat generation figure of merit, F 68 Watts/cm3 Average core power density < 124 Watts/cm3
ULNRC-06079 February 5, 2014 Page 4 of 5 To meet the upper axial boundary limitations, it is only necessary that the average core power density be less than 124 watts/cm3 and the distance between the top of the active fuel and the upper core plate be greater than 12.2 inches. The lower axial boundary criteria of MRP-227-A, Section 2.4, criteria, are satisfied by meeting the requirements stated above for average core power density, heat generation figure of merit, and the distance from the active fuel to the upper core plate. Thus, for the lower axial boundary criteria it is only necessary to meet the radial boundary and upper axial boundary limitations.
Historically, in-out core loading patterns have been used in all Callaway reload fuel cycles.
The average core power density has been 111.7 watts/cm3 since Cycle 3, when reactor power was uprated, and was 106.9 watts/cm3 prior to the uprating. Thus, the average core power density has always been met at Callaway.
With regard to the heat generation figure of merit, all reload fuel cycles except fuel cycles 2 and 13 met the limit of 68 watts/cm3. The duration of fuel cycle 2, which ran from April, 1986 to September, 1987, was 1.15 effective full power years. The duration of fuel cycle 13, which ran from November, 2002 to April, 2004, was 1.26 effective full power years. Although the heat generation figure of merit exceeded 68 watts/cm3 for these two fuel cycles in the first 20 years of operation, it did not exceed this value in the next 10 years of operation, and is not expected to exceed it in the second 30 years of operation. Since these two fuel cycles occurred in the first 20 years of operation, they do not invalidate the requirement to not use out-in loading patterns in the second 30 years of operation. In addition, the relatively short duration of these two fuel cycles in the first 30 years of operation are offset by many more years of operation where the heat generation figure of merit was below the limit.
The upper axial boundary criteria have always been met at Callaway. As discussed above, the average core power density is less than 124 watts/cm3. The distance from the active fuel to the upper core plate has varied due to changes in fuel design. However, this distance has always been greater than 12.2 inches, which meets the limit set by MRP 2013-025.
To ensure that these limits are met in future core designs, Callaway will continue to use in-out core loading patterns in all future fuel cycles. The core design procedure will be modified to include for each core loading pattern a review for the following parameters:
- Active fuel - upper core plate distance > 12.2 inches
- Average core power density < 124 watts/cm3
- Heat generation figure of merit, F 68 watts/cm3 LRA Table A4-1 item 43 has been added as shown in Amendment 29 in Enclosure 2 to revise the core design procedure to include the core design parameters noted above.
ULNRC-06079 February 5, 2014 Page 5 of 5 Corresponding Amendment Changes For Part (a) of the response, refer to the Enclosure 2 Summary Table "Amendment 30, LRA Changes for a description of LRA changes with this response.
For Part (b) of the response refer to the Enclosure 2 Summary Table "Amendment 29, LRA Changes for a description of LRA changes with this response (Refer to ULNRC-06072 dated January 16, 2014).
ULNRC-06079 February 5, 2014 Page 1 of 2 Callaway Plant Unit 1 License Renewal Application Amendment 30 Amendment 30, LRA Changes Summary Table Affected LRA Section LRA As-Submitted Page Number(s)
Table A4-1, item 44 A-49
ULNRC-06079 February 5, 2014 Page 2 of 2 Callaway Plant Unit 1 License Renewal Application Amendment 30 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item #
Commitment LRA Section Implementation Schedule 44 For all MRP-191 Table 4-4 components, as applicable to Callaway, Ameren Missouri commits to perform one or more of the following resolution options for the non-CASS RVI components:
Option 1: Replacement RVI components determined to be subject to 20% or greater cold work and 30 ksi operating stress will be replaced.
Option 2: Inspection For RVI components determined to be subject to 20% or greater cold work and 30 ksi operating stress, an augmented inspection program capable of detecting cracking will be developed. Minimum examination coverage criteria consistent with MRP-227-A Primary Inspection Category Components will apply. The augmented inspection program will be submitted to the NRC prior to performance of the inspection(s).
Option 3: Impact Evaluation For RVI components determined to be subject to 20% or greater cold work and 30 ksi operating stress, an impact evaluation will be prepared to establish that the effects of aging are minimal and will not have an adverse impact on future plant operability or component intended function. The impact evaluation(s) will be submitted to the NRC.
Option 4: Mitigation RVI components determined to be subject to 20% or greater cold work and 30 ksi operating stress will be mitigated of stress corrosion cracking (SCC) susceptibility.
Note: Indeterminate components will be conservatively assumed to be subject to 20% or greater cold work and subject to 30 ksi operating stress.
B2.1.6 To be completed no later than 24 months prior to the period of extended operation.
Page A-49