ML14126A434

Request for Additional Information for the Review of the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application, Set 24 (TAC Nos. MF1879, MF1880, MF1881, MF1882)

ML14126A434
Person / Time
Site:	Byron, Braidwood
Issue date:	05/19/2014
From:	Robinson L License Renewal Projects Branch 1
To:	Gallagher M Exelon Generation Co
Robinson L, 415-4115
References
TAC MF1880, TAC MF1881, TAC MF1882, TACE MF1879
Download: ML14126A434 (11)

Text

May 19, 2014 Mr. Michael P. Gallagher Vice President, License Renewal Projects Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE BYRON STATION, UNITS 1 AND 2, AND BRAIDWOOD STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION, SET 24 (TAC NOS. MF1879, MF1880, MF1881, AND MF1882)

Dear Mr. Gallagher:

By letter dated May 29, 2013, Exelon Generation Company, LLC, submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54, to renew the operating licenses NPF-37, NPF-66, NPF-72, and NPF-77 for Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, respectively, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.

These requests for additional information were discussed with John Hufnagel, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-4115 or e-mail Lindsay.Robinson@nrc.gov.

Sincerely,

/RA/

Lindsay R. Robinson, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-454, 50-455, 50-456, and 50-457

Enclosure:

Request for Additional Information cc: Listserv

ML14126A434 *concurred via email OFFICE LA:DLR* PM: RPB1:DLR BC:RPB1:DLR PM:RPB1:DLR NAME YEdmonds LRobinson YDiazSanabria LRobinson DATE 5/13/14 5/14/14 5/15/14 5/19/14

Letter to M.P. Gallagher from Lindsay R. Robinson dated May 19, 2014

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE BYRON STATION, UNITS 1 AND 2, AND BRAIDWOOD STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION, SET 24 (TAC NOS. MF1879, MF1880, MF1881, AND MF1882)

DISTRIBUTION EMAIL:

PUBLIC RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRarb Resource RidsNrrDlrRasb Resource RidsOgcMailCenter RidsNrrPMByron Resource RidsNrrPMBraidwood Resource

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LRobinson DMcIntyre, OPA JMcGhee, RIII EDuncan, RIII JBenjamin, RIII AGarmoe, RIII JRobbins, RIII VMitlyng, RIII PChandrathil, RIII

BYRON STATION, UNITS 1 AND 2, AND BRAIDWOOD STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION REQUEST FOR ADDITIONAL INFORMATION, SET 24 (TAC NOS. MF1879, MF1880, MF1881, MF1882)

RAI 3.5.2.2.1-1 Applicability:

Byron Station (Byron) and Braidwood Station (Braidwood), all units.

Background:

Subsequent to the issuance of Revision 2 of the Generic Aging Lessons Learned (GALL) Report (December 2010), NRC Information Notice (IN) 2011-20, Concrete Degradation by Alkali-Silica Reaction (ASR), was issued on November 01, 2011, to inform industry of operating experience related to concrete degradation due to ASR in Seismic Category 1 structures at Seabrook Station. IN 2011-20 states that American Society for Testing and Materials (ASTM) updated standards ASTM C1260 and ASTM C1293 and guidance provided in the appendices of ASTM C289 and ASTM C1293 caution that the tests described in ASTM C227 and ASTM C289 may not accurately predict aggregate reactivity when dealing with late-expanding or slow-expanding aggregates containing strained quartz or microcrystalline quartz. Therefore, licensees that tested using ASTM C227 and ASTM C289 could have concrete that is susceptible to ASR-induced degradation. Further, presence of ASR-induced degradation can be positively confirmed or refuted only by optical microscopy performed as part of petrographic examination of concrete core samples. As stated in NRC Inspection Report 05000443/2012010, dated August 9, 2013 (ADAMS Accession No. ML13221A172), the two root causes related to the occurrence of ASR degradation at Seabrook Station were: (1) the concrete mix design unknowingly utilized a coarse aggregate that would, in the long term, contribute to ASR and (2) the long-standing organizational belief that ASR was not a credible degradation mechanism due to the concrete mix design meeting industry standards and reactivity testing at the time of construction.

Section 3.5 in NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), includes several subsections (e.g.,

3.5.2.2.1.8, 3.5.2.2.2.1.2, 3.5.2.2.2.3.2) which identify the aging effect and mechanism of concrete cracking due to expansion from reaction with aggregates that do not require additional plant-specific aging management for inaccessible concrete areas if certain conditions identified in the GALL Report can be met. Table 3.5-1 in the SRP-LR includes line items for this aging effect and mechanism for accessible concrete areas in any environment and recommends the GALL Report aging management programs (AMPs) (e.g., XI.S2, ASME Section XI, Subsection IWL, and XI.S6, Structures Monitoring) to manage the effects of aging, with no associated conditions that can be met to consider the aging effect not applicable. The parameters monitored program element of the ASME Section XI, Subsection IWL and the Structures Monitoring AMPs in the GALL Report include this aging effect and mechanism. The NRC staff expects this aging effect and mechanism to be included within the recommended structural AMP.

ENCLOSURE

Issue:

License renewal application (LRA) Table 3.5.1, items 3.5.1-12, 3.5.1-19, 3.5.1-43, 3.5.1-50, and 3.5.1-54; and LRA Sections 3.5.2.2.1.8, 3.5.2.2.2.1.2, and 3.5.2.2.2.3.2, address cracking from expansion due to reaction with aggregates in concrete elements in accessible and inaccessible areas. The Discussion column of the aging management review (AMR) line items noted above states that this aging effect and mechanism does not apply to the respective Byron and Braidwood concrete structures because the fine and coarse aggregates used conform to ASTM C33, petrographic examinations of aggregates were performed in accordance with ASTM C295 and ASTM C289, and the concrete structures were constructed in accordance with American Concrete Institute (ACI) 318. The staff does not agree, since adequate plant-specific technical basis to support that statement has not been provided and the tests described in ASTM C227 and ASTM C289 may not accurately predict aggregate reactivity when dealing with late-expanding or slow-expanding aggregates containing strained quartz or microcrystalline quartz.

In light of the industry operating experience at Seabrook Station, unless positively justified, the staff position is that cracking due to expansion from reaction with aggregates for concrete in accessible and inaccessible areas could occur and should be managed through the period of extended operation. The discussion in the LRA sections referenced above states that although cracking associated with expansion due to reaction with aggregates has not been observed on Byron and Braidwood concrete structures, the respective structural AMPs mentioned, therein, will continue to inspect and monitor concrete structures for cracking due to any mechanism.

However, the associated line items for concrete cracking from expansion due to reaction with aggregates do not appear in any of the LRA Table 2s.

Request:

Provide technical justification why cracking due to expansion from reaction with aggregates (i.e.,

alkali-aggregate reaction) does not require management for concrete in accessible and inaccessible areas or identify applicable program(s) to manage this aging effect. If a program is identified to manage this aging effect, update the applicable LRA sections accordingly.

RAI 3.5.2-6 Applicability:

Byron and Braidwood

Background:

Item 24 located in SRP-LR Table 3.5-1 references the GALL Report item II.A1.CP-100. The AMP recommended for item II.A1.CP-100 in the GALL Report is XI.S2, ASME Section XI, Subsection IWL, or XI.S6, Structures Monitoring.

The GALL Report AMP XI.S2, ASME Section XI, Subsection IWL, Program Description states that 10 CFR 50.55a imposes the examination requirements of ASME Code,Section XI, Subsection IWL, for Class CC reinforced and prestressed concrete containments. The GALL Report AMP Scope of Program (Program Element 1) states that the components within the scope of Subsection IWL are reinforced concrete and unbonded post-tensioning systems of Class CC containments. Subsection IWL exempts from examination portions of the concrete containment that are inaccessible such as concrete covered by liner, foundation material, or backfill or obstructed by adjacent structures or other components. However,

10 CFR 50.55a(b)(2)(viii) specifies additional requirements for inaccessible areas that requires the licensee to evaluate the acceptability of concrete in inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation of inaccessible areas.

Issue:

The corresponding LRA Table 3.5.1, Item 3.5.1-24, states in the Discussion column that this item is consistent with the GALL Report and cites the Structures Monitoring (LRA Section B.2.1.34) program as the AMP for managing this aging effect and mechanism for inaccessible areas of containment concrete exposed to groundwater and soil environments, including groundwater chemistry. In the context of the LRA, the description for Item II.A1.CP-100 in the GALL Report does include containment pressure-resisting boundary concrete components in inaccessible areas above grade in an Air - Outdoor environment as well as below-grade areas in a Groundwater/Soil environment. The applicant has not addressed inaccessible components in the Air - Outdoor environment for this AMR line item. Further, the ASME Section XI, Subsection IWL program, mandated by the GALL Report and 10 CFR 50.55a for concrete containment pressure-resting boundary components in both accessible and inaccessible areas, is not included as an applicable AMP for AMR line item 3.5.1-24 and the corresponding line items in LRA Table 3.5.2-4.

Request:

With regard to AMR line Item 3.5.1-24 in LRA Table 3.5-1 that corresponds to Item II.A1.CP-100 in the GALL Report, provide the technical basis to justify why the ASME Code,Section XI, Subsection IWL program, recommended by the GALL Report and required by 10 CFR 50.55a for concrete containment pressure-resisting boundary components in both accessible and inaccessible areas, is not included as an applicable AMP for the line item and corresponding line items in LRA Table 3.5.2-4. Update the LRA, as necessary, based on the response to this request.

RAI B.2.1.12-1a Applicability:

Byron and Braidwood

Background:

The response to RAI B.2.1.12-1, dated February 27, 2014, stated that existing station procedures require a general visual inspection of internal surfaces of components within the scope of the Closed Treated Water Systems program when the systems are opened. In addition, the personnel performing the inspections are qualified to Exelon job qualifications and in accordance with the Institute of Nuclear Power Operations (INPO) National Academy for Nuclear Training accredited training program.

The staff notes that, similar to the Closed Treated Water Systems program, the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program uses opportunistic visual inspections to monitor aging effects of component internal surfaces. However, during its AMP audit, the staff noted that the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program does not rely on the use of existing station procedures to identify

age-related degradation. Rather, a new procedure was proposed to inspect for evidence of loss of material, leakage, cracking, and reduction of heat transfer when the internal surfaces of metallic components were made accessible.

Issue:

It is unclear to the staff how the existing station opportunistic inspections will be capable of detecting the specific applicable aging effects of components internal surfaces in the Closed Treated Water Systems program. The RAI response did not provide sufficient information regarding:

1. The details within the INPO training program and Exelon job qualifications that demonstrate that personnel performing the opportunistic inspections are qualified to identify the applicable aging effects, and

2. The details within the INPO training program, Exelon job qualifications, or existing station procedures that demonstrate that, when piping internal surfaces are made accessible, personnel will be inspecting for parameters that are capable of detecting the presence and extent of aging effects.

Request:

1. State the details within the INPO training program and Exelon job qualifications that demonstrate that personnel performing the opportunistic inspections are qualified to identify loss of material due to general, pitting, crevice, and galvanic corrosion; and cracking due to stress corrosion cracking.

2. State the details within the INPO training program and Exelon job qualification that will ensure that, when component internal surfaces are made accessible, personnel will be inspecting for parameters that are capable of detecting the presence and extent of loss of material due to general, pitting, crevice, and galvanic corrosion; and cracking due to stress corrosion cracking. Alternatively, state the process-based controls in existing station procedures (e.g., prejob brief details, checklists within the work order) that will ensure that personnel will be inspecting for the appropriate parameters.

RAI B.2.1.3-4 Applicability:

Braidwood

Background:

During the audit of the operating experience program element for Braidwood Units 1 and 2, the staff found that operating experience provided by the applicant in the LRA was incomplete.

Specifically, the applicants onsite database contained information related to a stuck reactor vessel closure stud at the Braidwood Unit 2. Based on the information provided by the applicant during the audit, Stud No. 35 became stuck during the 1991 outage, but it had enough thread engagement to be tensioned. The applicant was able to tension the stuck stud. The stuck stud was cut at the flange level in May of 1995 in order to facilitate safer fuel transfer during refueling outages. In an effort to repair Stud Hole No. 35, the remnant of the stud was bored out in 2002.

Due to human error, the stud bore hole was over bored, and the applicant decided to abandon its repair efforts. Currently, Braidwood Unit 2 has only 53 of 54 studs operable.

Issue:

The LRA does not provide any information regarding the significant plant-specific operating experience relative to Stud No. 35 for Braidwood Unit 2. In addition, no information was provided in the LRA or during the audit on the root cause of the failure. Without a root cause, the staff is concerned that similar failures could reoccur and further challenge the integrity of the reactor vessel head.

Request:

1. Perform a comprehensive plant-specific operating experience search for Braidwood Units 1 and 2. In addition to Stud No. 35, provide search results that include all instances of stuck studs, missing threads, damaged threads, or any form of degradation in reactor pressure vessel studs, guide studs, washers, vessel flange threads, and nuts.

2. Provide a detailed chronology of the events related to Braidwood Unit 2, Stud No. 35.

3. Provide a root cause analysis related to the failure of Stud No. 35. Include corrective actions, inspection results, engineering changes, repair replacement activities related to Stud No. 35 and its respective flange hole.

4. Provide details of the current configuration of Stud Hole No. 35 and inspection results from 2002 to present.

5. Provide inspection results for Stud and Stud Hole Nos. 33, 34, 36 and 37 for Braidwood Unit No. 2 from 1995 to present.

RAI B.2.1.24-1 Applicability:

Braidwood

Background:

In LRA Section B.2.1.24, the applicant provided brief discussions covering the operating experience of the Byron and Braidwood Units 1 and 2. In these discussions, it was noted that the Braidwood Unit 1 and 2 flux thimbles have experienced more wear than the Byron Unit 1 and 2 flux thimbles. Due to the observed higher wear rates, the examination frequency for both Braidwood Units was changed to every refueling outage. In addition, the operating experience provided in the LRA indicated that there have been instances when, either due to an obstruction or due to other outage related work, all the Braidwood flux thimbles were not examined.

Furthermore, the staffs review of operating experience data base for Braidwood also revealed that eddy current examinations were not performed for certain flux thimbles, due to the presence of moisture in the flux thimble tubes.

Issue:

The applicants plant-specific operating experience discussion in the LRA section states that Braidwood Units 1 and 2 flux thimble tubes examination frequency is every outage, due to higher than anticipated wear rates. The staff noted that the applicants operating experience discussion in the LRA did not fully address the reasons for the unexpected high wear rates observed for Braidwood Units 1 and 2 address all the issues during eddy current testing which precluded the testing of all flux thimbles. The staff is concerned about the sufficiency of the proposed AMP if these issues are not properly addressed and corrected.

Request:

1) Provide information in terms of root cause analyses and corrective actions which can explain and account for the higher than anticipated observed wear rates for Braidwood Units 1 and 2 flux thimble tubes.

2) Explain what root cause analyses and corrective actions have been performed to correct the occurrences of moisture in the thimble tubes given that these occurrences interfere in eddy current examinations of the flux thimble tubes.

3) Justify the adequacy of the program if the unexpected high wear rates are not accounted for and mitigated, given that there are issues related to the eddy current examinations of all flux thimble tubes (i.e., conflicting outage schedule, tube blockage, and the presence of moisture in the flux thimbles).

RAI 3.5.1-1 Applicability:

Byron and Braidwood

Background:

The LRA states, in Sections B.2.1.34 and B.2.1.30, respectively, that the Structures Monitoring and ASME Section XI, Subsection IWL programs are consistent, with enhancements, with the GALL Report Chapters XI.S6, Structures Monitoring, and XI.S2, ASME Section XI, Subsection IWL. The GALL Report XI.S6 Scope of Program program element states that the program includes all structures, structural components, component supports, and structural commodities in the scope of license renewal that are not covered by other structural AMPs (i.e., ASME Section XI, Subsection IWE (AMP X1.S1) and ASME Section XI, Subsection IWL (AMP X1.S2)).

The Parameters Monitored or Inspected program element states: If necessary for managing settlement and erosion of porous concrete sub-foundations, the continued functionality of a site de-watering system is monitored.

The GALL Report AMP XI.S2 Program Description section states that 10 CFR 50.55a imposes the examination requirements of ASME Code,Section XI, Subsection IWL, for Class CC reinforced and prestressed concrete containments. The Scope of Program program element states that the components within the scope of Subsection IWL are reinforced concrete and unbonded post-tensioning systems of Class CC containments. Subsection IWL exempts from examination portions of the concrete containment that are inaccessible such as concrete covered by liner, foundation material, or backfill or obstructed by adjacent structures or other

components. However, 10 CFR 50.55a(b)(2)(viii) specifies additional requirements for inaccessible areas that require the licensee to evaluate the acceptability of concrete in inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation of inaccessible areas.

LRA Table 3.5.1, Item 3.5.1-1 corresponds to Item II.A1.CP-101 in the GALL Report for containment concrete components (Concrete: dome; wall; basemat; ring girders; buttresses; Concrete elements, all) for the aging effect of concrete cracking and distortion due to the aging mechanism of increased stress levels from settlement in a soil environment. For this item, the LRA states in the Discussion column: Consistent with NUREG-1801. The Structures Monitoring (B.2.1.34) program will be used to manage cracking and distortion of the concrete dome, wall, basemat, and buttresses in inaccessible areas of the Containment Structure exposed to a groundwater and soil environment. [Byron and Braidwood Stations] BBS do not rely upon a de-watering system to control settlement. See subsection 3.5.2.2.1.1. LRA Section 3.5.2.2.1.1 states that Item 3.5.1-1 is applicable to Byron and Braidwood and that inaccessible below grade containment concrete surfaces will be examined by the Structures Monitoring (B.2.1.34) program when excavated for any reason.

Issue:

For the aging effect of cracking and distortion due to increase in stress levels from settlement, although settlement can occur in a soil environment, the symptoms can be manifested in either an air-indoor uncontrolled or air-outdoor environment (see Table IX.E of the GALL Report for the aging effect term Cracks; distortion; increase in component stress level). NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), Section 3.5.2.2.1.1 acceptance criterion for the aging effect or mechanism of cracking and distortion due to increased stress levels from settlement states that the existing program relies on ASME Section XI, Subsection IWL to manage these aging effects on concrete components of the containment pressure-resisting boundary in both accessible and inaccessible areas.

Contrary to the scope and program descriptions for the GALL Report AMP XI.S2 and GALL Report AMP XI.S6 and the SRP-LR Section 3.5.2.2.1.1 acceptance criterion, LRA Table 3.5.1-1 does not identify the ASME Section XI, Subsection IWL as an applicable AMP. Also, LRA Table 3.5.2-4 does not identify the aging effect or mechanism corresponding to LRA Table 3.5-1, Item 3.5.1-1, for containment pressure boundary concrete component (Concrete: dome; wall; basemat; ring girders; buttresses; reinforcing steel) as an aging effect requiring management for accessible areas and for inaccessible areas; it identifies the Structures Monitoring Program as the only applicable AMP. The applicant has not provided the technical basis justifying these determinations and consistency with NUREG-1801 in the LRA for LRA Table 3.5.1, Item 3.5.1-1, and corresponding line items in LRA Table 3.5.2-4.

Request:

With regard to AMR line item 3.5.1-1 in LRA Table 3.5-1 that corresponds to Item II.A1.CP-101 in the GALL Report, provide the technical basis to justify: (a) why the ASME Section XI, Subsection IWL program is not listed for aging management of concrete containment pressure-resisting boundary components in accessible and inaccessible areas for this AMR line item and corresponding items in LRA Table 3.5.2-4 and (b) why the aging effect or mechanism

corresponding to the AMR line item is not identified as an aging effect or mechanism requiring management in LRA Table 3.5.2-4 for containment pressure boundary concrete components in accessible areas.