View source for ULNRC-06090, Enclosure 1: Callaway Plant Unit 1, Response to Request for Additional Information Set 30 and Updated Response to RAI B2.1.6-4d, Part (B)

{{Adams
| number = ML14073A004
| issue date = 03/13/2014
| title = Enclosure 1: Callaway Plant Unit 1, Response to Request for Additional Information Set 30 and Updated Response to RAI B2.1.6-4d, Part (B)
| author name = 
| author affiliation = Ameren Missouri, Union Electric Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000483
| license number = NPF-030
| contact person = 
| case reference number = ULNRC-06090
| package number = ML14073A002
| document type = - No Document Type Applies
| page count = 7
}}

=Text=
{{#Wiki_filter:ULNRC-06090 March 13, 2014    Page 1 of 7   

CALLAWAY PLANT UNIT 1 LICENSE RENEWAL APPLICATION 

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION (RAI) SET # 30 AND UPDATED RESPONSE TO RAI B2.1.6-4d, PART (b) 

ULNRC-06090 March 13, 2014    Page 2 of 7  RAI 3.1.2.1-6 

==Background:==
By letter dated October 24, 2012, Union Electric Company (Ameren Missouri) provided License Renewal Application (LRA) Amendment No. 13 which revised LRA Table 3.1.2-1, "Reactor Vessel, Internals, and Reactor Coolant System -Summary of Aging Management Evaluation - Reactor Vessel and Internals," to provide an amended aging management review (AMR) for the reactor vessel internals clevis insert bolts. In this amended AMR, the applicant confirmed that the clevis insert bolts are fabricated from nickel alloy materials and that potential cracking and loss of material due to wear will be managed by the applicant's American Society of Mechanical Engineers (ASME) Section XI Inservice Inspection (lSI), Subsections IWB, IWC, and IWD Program. Issue: Appendix A to Electric Power Research Institute (EPRI) Technical Report No. 1022863, "Materials Reliability Program: Pressured Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (MRP-227-A)," indicates that failures of nickel alloy (Alloy X-750) clevis insert bolts were reported by the licensee for one domestic Westinghouse-designed pressurized water reactor in 2010. The ASME Section XI visual VT-3 inspections of the clevis insert assemblies on a 10-year frequency may not be adequate to ensure the integrity of clevis insert assemblies during design basis events if multiple bolt failures occur prior to detection and a design basis event occurs. 

Request:  
(a)    Describe the configuration of clevis insert assemblies at Callaway Plant, Unit 1 (Callaway)              including number of bolts in the assemblies. Specify the material of fabrication, including              any applicable heat treatment, that were used for the design of the clevis insert bolts at              Callaway. 

(b)  Discuss and justify whether the operating experience associated with cracking of the clevis              insert bolts is applicable to clevis insert assembly designs at Callaway. 

(c) Describe the inspections that have been performed of the clevis insert bolts, including the              type of inspection (e.g., VT-3). Clarify the visual inspection coverage that was achieved during these inspections. Clarify the ASME examination category that applies to inspections of the clevis insert bolts (and identify the applicable inspection method and frequency) and whether the past examinations have resulted in the detection of any indications of cracking or failures of the clevis insert bolts that are included in the clevis insert assembly designs. If so, provide the details of the inspection results and clarify the corrective actions that were taken at the facility to justify the structural integrity of the clevis insert assemblies and the intended safety function of the plant's core support structure and its components during plant operations. 

(d) Based on your responses to Parts (a) through (c) of this request, clarify whether the 10-year lSI basis for the clevis insert bolts is sufficient to manage cracking and wear of the bolts during the period of extended operation. Justify your response to this request. 

ULNRC-06090 March 13, 2014    Page 3 of 7  Callaway Response (a) The clevis insert assemblies at Callaway are comprised of eight (8) clevis insert bolts fabricated from Inconel X-750, two (2) dowel pins and an insert assembly that rests onto the clevis locations in the reactor vessel. In total there are six (6) clevis locations (0, 60, 120, 180, 240, 300 degree locations). The clevis insert bolts include a lock bar that is welded to the clevis insert face. The clevis insert bolts are attached to the interior face of the clevis insert assembly. 

The Callaway material used for the clevis insert bolts is the same as used at the reference plant where cracking has been observed. The clevis insert design however is different than the reference plant in that the clevis insert bolt locations are included on the interior face of the clevis (parallel to the outside face of the radial insert key with the core barrel installed). This geometry is further depicted in the Clevis Assembly Diagram, as shown below.   

Heat treatment information will be provided in a later submittal. 
 (b) The response to part (b) will be provided in a later submittal. 
 (c) Callaway performs VT-3 inspections on 100% of accessible components in accordance with ASME Section XI B-N-2 code categorization. Callaway has performed this inspection on multiple occasions. The most recent inspection was performed in Spring 2013 during Callaway's Refuel 19, in which 100% of the clevis insert bolt heads and lock bars were observed in conjunction with the B-N-2 inspection and no degradation or damage was identified. There were no inspection results that identified similar clevis insert bolt head detachment, as was identified at the reference plant. 

Based on these considerations and inspection results, the Callaway Reactor Vessel Internals Program will not be augmented for crack detection of the lower radial support clevis insert bolts. 

Callaway continues to monitor operating experience in this area and will review applicable operating experience for program modifications.  (d) The response to part (d) will be provided in a later submittal.
Corresponding Amendment Changes No changes to the License Renewal Application (LRA) are needed as a result of this response. 

ULNRC-06090 March 13, 2014    Page 4 of 7 ULNRC-06090 March 13, 2014    Page 5 of 7  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-06090 March 13, 2014    Page 6 of 7 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. A subsequent response to part (a) was transmitted by ULNRC-06079 date February 5, 2014.          
(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/cm 3 Average core power density < 124 Watts/cm 3     To meet the upper axial boundary limitations, it is only necessary that the average core power density be less than 124 watts/cm 3 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 ULNRC-06090 March 13, 2014    Page 7 of 7  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/cm 3 since Cycle 3, when reactor power was uprated, and was 106.9 watts/cm 3 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/cm 3 for Type 1 corners. For Type 2 corners, fuel cycle 3 was the only reload cycle which did not meet the limit of 68 watts/cm 3. 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 3, which ran from November, 1987, to March, 1989, was 1.23 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/cm 3 for these two three 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 three 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 three 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/cm
: 3. 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/cm 3
* Heat generation figure of merit, F  68 watts/cm 3
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. This commitment was subsequently closed as indicated in LRA Amendment 31 transmitted by ULNRC-06080 dated February 14, 2014 (Reference 5).  

Corresponding Amendment Changes Refer to the Enclosure 2 Summary Table "Amendment 29, LRA Changes" for a description of LRA changes with this response.}}