ML24097A007

Relief Request Number RR-ENG-4-07 – Request for an Alternative to ASME Code Case N-729-6 for Reactor Vessel Head Penetration 75

ML24097A007
Person / Time
Site:	South Texas
Issue date:	04/06/2024
From:	Georgeson C South Texas
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NOC-AE-24004031, 35582628
Download: ML24097A007 (1)

Text

April 6, 2024 NOC-AE-24004031 STI: 35582628 10 CFR 50.55a

Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

South Texas Project Unit 2 Docket No. STN 50-499 Relief Request Number RR-ENG -4 Request for an Alternative to ASME Code Case N-729-6 For Reactor Vessel Head Penetration 75

In accordance with the provisions of 10 CFR 50.55a(z)(2), STP Nuclear Operating Company (STPNOC) hereby requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Code Case N-729-6, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial -Penetration WeldsSection XI, Division 1," for use at South Texas Project Unit 2.

On April 1, 2024, during refueling outage 2RE23, STPNOC identified evidence of leakage on the Unit 2 Reactor Vessel Head (RVH) near Core Exit Thermocouple Nozzle Assembly (CETNA) penetration 75 during a visual examination (VE). The indication was boron crystals and discoloration with corrosion products present that is indicative of a Reactor Coolant System (RCS) leak.

STPNOC is requesting relief from Code Case N -729-6 paragraphs -3142.2 and -3200 for performing volumetric and/or surface exams of the STP Unit 2 RVH CETNA penetration 75. As an alternative, STPNOC proposes to perform a bare metal VE of CETNA penetration 75 at the next refueling outage after cleaning and documentation of the as left condition ( 2RE24, Fall 2025). The proposed alternative and supporting information are presented in the Enclosure.

STPNOC requests approval of the proposed alternative by April 9, 2024.

There are no commitments in this letter.

If there are any questions regarding this matter, please contact Drew Richards at (361) 972-7666 or me at (361) 972-7806.

C. H. Georgeson General Manager, Engineering

Enclosure:

10 CFR 50.55a Request for an Alternative to ASME Code Case N -729-6 for Reactor Vessel Head CETNA Penetration 75

NOC-AE-24004031 Page 2 of 2 cc:

USNRC Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 E. Lamar Boulevard Arlington, TX 76011-4511

NOC-AE-24004031 Enclosure

ENCLOSURE

10 CFR 50.55a Request for an Alternative to ASME Code Case N -729 -6 for Reactor Vessel Head CETNA Penetration 75

NOC-AE-24004031 Enclosure Page 1 of 7

Enclosure

10 CFR 50.55a Request for an Alternative to ASME Code Case N -729 -6 for Reactor Vessel Head CETNA Penetration 75

1. ASME CODE COMPONENT(S) AFFECTED:

Component: Replacement Reactor Vessel Head (RVH) nozzles Code Class: Class 1 Exam Category: ASME Code Case N-729-6, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure - Retaining Partial-Penetration Welds,Section XI, Division 1 Code Item No.: B4.30

Description:

CORE EXIT THERMOCOUPLE NOZZLE ASSEMBLY (CETNA), Nozzle 75 Size: Ø4.000" OD x Ø2.750" ID x 0.625" wall Materials: RVH - SA-508 Class 3 Nozzles - SB-167 N06690 (Alloy 690)

Weld Material: ERNiCrFe-7, ENiCrFe -7 (Alloy 690 weld material)

There are f ifty-seven Alloy 690 Control Rod Drive Mechanism ( CRDM) head penetration nozzles, four spare capped Alloy 690 CRDM head penetration, four CETNA core exit thermocouple head penetration nozzles, two Reactor Vessel Water Level Indicating System head penetration nozzles, and one head vent welded to the inside surface of the RVH with partial penetration J-groove welds.

2. APPLICABLE CODE EDITION AND ADDENDA:

The Fourth Ten-Year ISI interval Code of record for South Texas Project (STP) Unit 2 is the 2013 Edition ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Examinations of the reactor vessel head ( RVH) penetrations are performed in accordance with ASME Code Case N -729-6 ( Reference 1) as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) (2) through (8).

The manufacturing Code for STP Unit 2 RVH: ASME Boiler and Pressure Vessel (BPV) Code,Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, 1989 Edition.

3. APPLICABLE CODE REQUIREMENT:

The Code of Federal Regulations (CFR) 10 CFR 50.55a(g)(6)(ii)(D)(1), requires (in part):

Augmented ISI requirements: Reactor vessel head inspections (1) Implementation. Holders of operating licenses or combined licenses for pressurized-water reactors as of or after June 3, 2020 shall implement the requirements of ASME BPV Code Case N-729-6 instead of ASME BPV Code Case N-729-4, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (8) of this section, by no later than one year after June 3, 2020. All previous NRC-approved alternatives from the requirements of paragraph (g)(6)(ii)(D) of this section remain valid.

Code Case N-729-6, Paragraph -3141, Inservice Visual Examinations (VE) states:

(a) The VE required by -2500 and performed in accordance with IWA -2200 and the additional requirements of this Case shall be evaluated by comparing the examination results with the acceptance standards specified in -3142.1.

(b) Acceptance of components for continued service shall be in accordance with -3142.

NOC-AE-24004031 Enclosure Page 2 of 7

(c) Relevant conditions for the purposes of the VE shall include evidence of reactor coolant leakage, such as corrosion, boric acid deposits, and discoloration.

Code Case N-729-6, Paragraph 3142.1, Acceptance by VE states:

(a) A component whose VE confirms the absence of relevant conditions shall be acceptable for continued service.

(b) A component whose VE detects a relevant condition shall be unacceptable for continued service until the requirements of (1), (2), and (c) below are met.

(1) Components with relevant conditions require further evaluation. This evaluation shall include determination of the source of the leakage and correction of the source of leakage in accordance with -3142.3.

(2) All relevant conditions shall be evaluated to determine the extent, if any, of degradation. The boric acid crystals and residue shall be removed to the extent necessary to allow adequate examinations and evaluation of degradation, and a subsequent VE of the previously obscured surfaces shall be performed, prior to return to service, and again in the subsequent refueling outage. Any degradation detected shall be evaluated to determine if any corrosion has impacted the structural integrity of the component. Corrosion that has reduced component wall thickness below design limits shall be resolved through repair/replacement activity in accordance with IWA-4000.

(c) A nozzle whose VE indicates relevant conditions indicative of possible nozzle leakage shall be unacceptable for continued service unless it meets the requirements of -3142.2 or -3142.3.

Code Case N-729-6, Paragraph -3142.2, Acceptance by Supplemental Examination states:

A nozzle with relevant conditions indicative of possible nozzle leakage shall be acceptable for continued service if the results of supplemental examinations [-3200(b)]

meet the requirements of -3130.

Code Case N-729-6, Paragraph -3142.3, Acceptance by Corrective Measures or Repair/Replacement Activity states:

(a) A component with relevant conditions not indicative of possible nozzle leakage is acceptable for continued service if the source of the relevant condition is corrected by a repair/replacement activity or by corrective measures necessary to preclude degradation.

(b) A component with relevant conditions indicative of possible nozzle leakage shall be acceptable for continued service if a repair/replacement activity corrects the defect in accordance with IWA-4000.

Code Case N-729-6, Paragraph -3200, Supplemental Examination states:

(a) Volumetric or surface examinations that detect flaws which require evaluation in accordance with -3130 may be supplemented by other techniques to characterize the flaw (i.e., size, shape, and orientation).

(b) The supplemental examination performed to satisfy -3142.2 shall include volumetric examination of the nozzle tube and surface examination of the partial penetration weld, or surface examination of the nozzle tube inside surface, the partial penetration weld, and nozzle tube outside surface below the weld, in accordance with Figure 2, or the alternative examination area or volume shall be analyzed to be acceptable in accordance with Mandatory Appendix I. The supplemental examinations shall be used to determine the extent of the unacceptable conditions and the need for corrective measures, analytical evaluation, or repair/replacement activity.

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4. REASON FOR REQUEST:

STP Nuclear Operating Company (STPNOC) is requesting relief from Code Case N-729-6 paragraphs -3142.2 and -3200 for performing supplemental volumetric and/or surface exams of the STP Unit 2 RVH CETNA penetration 75 due to a relevant condition identified on the RVH surface adjacent to this penetration. Once a relevant condition of possible nozzle penetration leakage is identified, Code Case N-729-6, paragraph -3142.2 requires supplemental examination including a volumetric examination of the nozzle tube and surface examination of the partial penetration weld in accordance with paragraph -3200(b) that demonstrates the acceptance of the nozzle.

During the ongoing STP Unit 2 refueling outage 2RE23 (Spring 2024), STPNOC performed a scheduled VE of the RVH nozzle penetrations in accordance with Code Case N -729-6 (Reference 1). This was the third VE of the replacement RVH. During the examination, Engineering identified leakage around head penetration 75. Subsequent inspections identified the source of leakage was from the CETNA in penetration 75.

Leakage from the CETNA flange flowed down the CETNA tube to the tube-to-head penetration as evidenced by the alluvial trails that can be seen on the CETNA tube. The leakage pooled on the downhill side of the annulus area of penetration 75 and then continued to flow downhill as alluvial trails can be seen to continue further down on the head surface. Although the leakage caused staining with corrosion product deposits and discoloration, the appearance and pattern was not consistent with known operational boric acid leaks coming from RVH penetrations as documented in EPRI Report MRP-60 ( Reference 2). The leakage from the CETNA was not characteristic of leakage expected from a crack or defect.

STPNOC evaluated this condition using the guidance from 0POP01 -ZO -0011, Operability, Functionality and Reportability Guidance, Addendum 1, Pressure Boundary Leakage (Reference 3). The CETNA leakage was positively identified as being from a mechanical joint and is therefore not pressure boundary leakage. Additionally, previous e ngineering evaluation concluded the leakage from CETNA penetration 75 occurred during pressurization and heat-up from 2RE22 in Fall 2022, not during normal operation. Therefore, c ontinued active leakage is not anticipated during the cycle.

Performing the Supplemental Examinations in accordance with Code Case N -729-6, paragraph -3200(b) requires access to the underside of the irradiated RVH which would expose personnel to elevated dose rates. Dose rates under the RVH are estimated to be between 650 mR/hr and 1500 mR/hr based on a recent historical survey for Unit 2. The total estimated personnel dose to perform the Supplemental Examination is estimated to be approximately 9.0 man-rem for performing this work activity, which includes setup, inspection, and demobilization. This dose impact could be significantly reduced given time to properly plan the activity and consider potential dose mitigation measures.

The relevant indication is not indicative of reactor vessel head leakage. Based on previous engineering evaluation, STPNOC has concluded there is reasonable assurance there will not be continued active leakage from the CETNA during the next operating cycle. Therefore, the significant increase to personnel dose that could result from performing the Supplemental Examinations required by Code Case N-729-6 paragraphs - 3142.2 and -3200 represents a hardship without a compensating increase in the level of quality and safety pursuant to 10 CFR 50.55a(z)(2). All other ASME Code,Section XI requirements for which an alternative was not specifically requested and approved remain applicable.

NOC-AE-24004031 Enclosure Page 4 of 7

5. PROPOSED ALTERNATIVE AND BASIS FOR USE:

STPNOC is requesting relief from Code Case N-729-6 paragraphs -3142.2 and -3200 for performing volumetric and/or surface exams of the STP Unit 2 RVH penetration 75. As an alternative, STPNOC proposes to perform a bare metal VE of CETNA penetration 75 at the next refueling outage (2RE24, Fall 2025) after cleaning and documentation of the as left condition.

Previous Examinations of the South Texas 2 Replacement Head The STP Unit 2 replacement head went into service in Spring 2010 at STP Unit 2 (approximately 14 years of service). Prior to installation, a preservice volumetric examination of the replacement RVH nozzles was performed. There were no recordable indications identified during the preservice volumetric examinations of the nozzle tube in the area of the J-groove welds.

A bare metal VE was performed of the STP Unit 2 replacement RVH in Spring 2015 in accordance with ASME Code Case N-729-1 ( Reference 6), Table 1, Item B4.30. This VE was performed by VT-2 qualified examiners, on the outer surface of the RVH including the annulus area of the penetration nozzles. This examination did not reveal any surface or nozzle penetration boric acid that would be indicative of nozzle leakage.

A bare metal VE was performed of the STP Unit 2 replacement RVH in SL2-23, Fall 2019 in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. This VE was performed by VT-2 qualified examiners on the outer surface of the RVH including the annulus area of the penetration nozzles. This examination did not reveal any surface or nozzle penetration boric acid that would be indicative of nozzle leakage.

Note: All the above visual examinations were performed by qualified VT-2 examiners performing VE exams per Code Case N-729-1 and later versions have documented additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of training in detection of borated water leakage per Reference 1 and Reference 6, Table 1 Note 2 as required by the STP non-destructive examination ( NDE) procedure.

MRP-375 Information Regarding the Structural Adequacy and Performance of the RVH Alloy 690 Nozzles Evaluations were performed and documented in EPRI MRP -375 (Reference 4) to demonstrate the acceptability of extending the RVH inspection intervals for ASME Code Case N-729-1, item B4.40 components based on the superior laboratory and operational performance.

Per MRP-375, much of the laboratory data indicated a Factor of Improvement (FOI ) of 100 for Alloy 690/52/152 versus Alloy 600/182/82 (for equivalent temperature and stress conditions) in terms of crack growth rates (CGR). In addition, laboratory and plant data demonstrate an FOI in excess of 20 in terms of the time to PWSCC initiation. This reduced susceptibility to PWSCC initiation and growth supports elimination of all volumetric examinations throughout the plant service period, and by extension, supports not performing Supplemental Examinations this refueling outage.

Alloy 690 is highly resistant to Primary Water Stress Corrosion Cracking (PWSCC) due to its approximate 30% chromium content. Per MRP -115 ( Reference 5), it was noted that Alloy 82 CGR is 2.6 times slower than Alloy 182. There is no strong evidence for a difference in Alloy 52 and 152 CGRs. Therefore, data used to develop factors of improvement for Alloy 52/152 were referenced against the base case Alloy 182, as Alloy 182 is more susceptible to initiation and growth when compared to Alloy 82. A simple factor of improvement (FOI) approach was applied in a conservative manner in MRP -375 using multiple data. Based on plant service experience, FOI studies using laboratory data, deterministic study results, and probabilistic study results, MRP-375 documented the basis for extended inspection intervals.

This information documents the structural suitability of the RVH for extended periods of time.

NOC-AE-24004031 Enclosure Page 5 of 7

Deterministic calculations demonstrate that the alternative volumetric re-examination schedule of MRP-375 (Table 4-1) of every 20 years is sufficient to detect any PWSCC before it could develop into a safety significant circumferential flaw that approaches the large size (i.e., more than 300 degrees of circumferential extent) necessary to produce a nozzle ejection with significant margins of safety. The deterministic calculations also demonstrate that any base metal PWSCC would likely be detected prior to a through-wall flaw occurring.

Probabilistic calculations based on a Monte Carlo simulation model of the PWSCC process, including PWSCC initiation, crack growth, and flaw detection via ultrasonic testing, show a substantially reduced effect on nuclear safety compared to a RVH with Alloy 600 nozzles examined per current requirements.

Service Experience As documented in MRP-375 (published in 2014), the resistance of Alloy 690 and corresponding weld metals Alloy 52 and 152 is demonstrated by the lack of any PWSCC indications reported in these materials, in up to 24 calendar years of service for thousands of Alloy 690 steam generator tubes, and more than 22 calendar years of service for thick -wall and thin-wall Alloy 690 applications. There has been no new operating experience of PWSCC identified since the publication. This excellent operating experience inc ludes service at pressurizer and hot-leg temperatures and includes Alloy 690 wrought base metal and Alloy 52/152 weld metal. This experience includes ISI volumetric or surface examinations performed in accordance with ASME Code Case N -729-1 on at least 13 of the 41 replacement RVHs currently operating in the U.S. fleet. This data supports a factor of improvement in time of at least 5 to 20 to detectable PWSCC when compared to service experience of Alloy 600 in similar applications.

The STP Unit 2 head was fabricated using thermally treated Alloy 690 nozzle material. The nozzle J-groove attachment welds for the STP Unit 2 head utilized PWSCC resistant ERNiCrFe-7 (UNS N06052 and/or ENiCrFe-7 UNS W86152) weld materials. The STP Unit 2 was procured to ASME Section III, 1989 Edition, no addenda.

Design Features Further Increasing the Resistance of the South Texas Unit 2 Replacement Head to PWSCC In addition to the standard Alloy 690 materials (plate and CRDM nozzle material) test data reported in MRP-375, STP imposed supplemental requirements on the STP Unit 2 nozzle materials to increase the material resistance to PWSCC. These supplemental requirements include thermal treatment (TT), additional chemistry requirements, microstructure and grain size requirements.

These methods substantially reduce PWSCC susceptibility beyond that assumed in the generic MRP-375 study, resulting in additional assurance that the STP Unit 2 head penetrations are highly resistant to PWSCC.

As stated above, none of the prior examinations of replacement RVHs and pressurizer with Alloy 690 nozzles have revealed any indications of PWSCC or service-induced cracking.

Enhance RCS Leakage Detection at South Texas Unit 2 Provides Defense in Depth As discussed above, the initiation or growth of a safety significant flaw in an alloy 690 base material and associated weld material in a RVH penetration is extremely unlikely.

However, as an added measure of safety, the industry imposed an NEI-03-08 needed requirement, to improve their RCS leak detection capability in part due to the concern with PWSCC or alloy 600 materials. STP Unit 2 has adopted the standardized approach to measuring RCS leak rate in WCAP-16423 ( Reference 7) and has incorporated the action levels in WCAP-16465 ( Reference 8). The enhanced leak rate monitoring and detection procedure monitors specific values of unidentified leakage, the seven-day rolling average, and the baseline means. Action levels are initiated as low as when the unidentified leak rate NOC-AE-24004031 Enclosure Page 6 of 7

exceeds 0.1 gpm. The enhanced leak detection capability provides an increased level of safety that if a flaw were to grow through wall, although unlikely, that is would be detected prior to it growing to a safety significant size.

Conclusion STPNOC has concluded that there is reasonable assurance that the relevant indication at the STP Unit 2 RVH penetration 75 is not indicative of RCS leakage from penetration 75 base material or partial penetration weld based on the following:

• The relevant condition was noted on the RVH surface adjacent to CETNA penetration 75.

Leakage source identified from the CETNA flange flowed down the CETNA tube to the tube-to-head penetration as evidenced by the alluvial trails that can be seen on the CETNA tube.

• The leakage flowed down the CETNA tube and pooled on the downhill side of the annulus area of penetration 75 and then continued to flow downhill as alluvial trails can be seen to continue further down on the head surface.

• Although the leakage evidence caused staining with corrosion product deposits and discoloration, the appearance and pattern was not consistent with known operational boric acid leaks coming from RVH penetrations as documented in EPRI Report MRP -60 (Reference 2).

In addition, the South Texas Unit 2 RVH has only been in service since 2010 with 14 years of operation. Operating experience and laboratory testing of Alloy 690 materials and the associated alloy 52/152 weld materials (ERNiCrFe-7 and/or ENiCrFe-7) show significant resistance to PWSCC with factors of improvement over alloy 600 materials and supports reinspection intervals of 20 years with margins of safety.

The relevant indication is not indicative of reactor vessel head leakage. Based on previous engineering evaluation, STPNOC has concluded there is reasonable assurance there will not be continued active leakage from the CETNA during the next operating cycle. Therefore, performing the Supplemental Examinations required by Code Case N -729-6 paragraphs -

3142.2 and -3200 that would subject workers to approximately 9.0 man-Rem of dose represents a hardship without a compensating increase in the level of quality and safety pursuant to 10 CFR 50.55a(z)(2).

Based on the discussion and the summary above, it is requested that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(2) as the alternative provides an acceptable level of quality and safety.

6. DURATION OF PROPOSED ALTERNATIVE:

The proposed alternative is for one operating cycle, until the end of operating cycle 2 4 (Fall 2025 ), at which time STPNOC will perform a bare metal VE of CETNA penetration 75 pursuant to Code Case N-729-6.

NOC-AE-24004031 Enclosure Page 7 of 7

7. PRECEDENTS:

1. NRC letter regarding verbal approval of Relief Request 19 for St. Lucie Nuclear Plant, Unit 2,

Subject:

St. Lucie Nuclear Plant, Unit 2 - Verbal Authorization of Relief Request 19, FP&L letter L-2021-184 (EPID L-2021-LLR -0065), dated September 15, 2021 (ADAMS Accession Number ML22011A085)

2. NRC letter regarding approval of Relief Request (RR) 14 for Fort Calhoun Station, Unit No. 1,

Subject:

Fort Calhoun Station, Unit No. 1 - Request for Relief RR-14, From Certain Requirements of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds, dated August 21, 2015 (ADAMS Accession number ML15232A003)

3. NRC letter regarding approval of Relief Request 57 for Palo Verde Generating Station, Unit 1 - Relief Request No. 57 To Approve Alternate Requirements For The Reactor Pressure Vessel Head Nozzles To Perform A Bare Metal Examination Per ASME Code Case N-729-4, dated February 20, 2018 ( ADAMS Accession number ML18040A331)

8.

REFERENCES:

1. ASME Code Case N-729-6, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial -Penetration Welds,Section XI, Division 1, Approved March 3, 2016.

2. Materials Reliability Program: Visual Examination for Leakage of PWR Reactor Vessel Upper Head Nozzles:(MRP-60, Rev 5). EPRI, Palo Alto, CA: 2018. 3002013268 (ML020090G363 - Transmittal of Proprietary MRP-60, Rev 5)

3. STP Procedure 0POP01-ZO -0011, Operability, Functionality, and Reportability Guidance, Revision 18, effective April 6, 2023

4. Materials Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP -375), EPRI, Palo Alto, CA: 2014. 3002002441. [freely available at www.epri.com]

5. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP -115), EPRI, Palo Alto, CA: 2004. 1006696. [freely available at www.epri.com]

6. ASME Code Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial -Penetration Welds,Section XI, Division 1, Approved March 28, 2006.

7. WCAP-16423-NP, Rev. 0, Pressurized Water Reactor Owners Group Standard Process and Methods for Calculating RCS Leak Rate for Pressurized Water Reactors, Westinghouse Electric Co., September 2006. (Transmitted to the NRC - ML070310081, ML070310084)

8. WCAP-16465-NP, Rev. 0, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors, Westinghouse Electric Co., September 2006. (Transmitted to the NRC - ML070310081, ML070310082)

9. ATTACHMENT:

Supporting Figures and Photos of As -Found and As -Left Conditions

NOC-AE-24004031 Attachment

Attachment

Supporting Figures and Photos of As-Found and As-Left Conditions NOC-AE-24004031 Attachment Page 1 of 3

Figure 1 - South Texas Project Unit 2 RVH Penetration Layout with Penetration 75 Identified

NOC-AE-24004031 Attachment Page 2 of 3

Figure 2 - Boron and Discoloration As-Found Condition at Penetration 75 Figure 3 - As-Left Condition at Penetration 75

Figure 5 - Boron Deposits As-Found Condition Above Penetration 75

Figure 4 - Boron Deposits Removed As-Left Above Penetration 75

NOC-AE-24004031 Attachment Page 3 of 3

Figure 7 - Boron and Discoloration As-Found Condition at Figure 6 - As-Left Condition at Penetration 75 Penetration 75

Figure 8 - Boron and Discoloration As-Found Condition at Figure 9 - As-Left Condition at Penetration 75 Penetration 75