ML18129A334

Corrosion Evaluation of the Limerick Unit 2 N-160 Reactor Vessel Nozzle Modification, Framatome Document No. 51-9271770-002 (Non-Proprietary)

ML18129A334
Person / Time
Site:	Limerick
Issue date:	05/04/2018
From:	Exelon Generation Co, Framatome
To:	Office of Nuclear Reactor Regulation
References
LG-18-068 51-9271770-002
Download: ML18129A334 (15)

Text

Attachment 8 "Corrosion Evaluation of the Limerick Unit 2 N-160 Reactor Vessel Nozzle Modification - Non-Proprietary," Framatome Document No. 51-9271770-002, Non-Proprietary Version

fr mt m Framatome Inc.

Engineering Information Record Document No.: - --9271770

- - -- -002 --

Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification.:.. Non-Proprietary Page 1 of 14

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framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Safety Related? ~ YES 0No Does this document ~tablish design 01* tecbnical reqtiirements? D YES !ZI NO Does this document contain assumptions requil'ing verification? 0 YES IZ!No Does this document contain Customer Required Format? DYES IZ! NO Signature Block Pages/Sections Name and P/LP, R/LR, M, Prepared/Reviewed/

Title/Discipline Signature 'A-CRF,A Date Approved or Comments RyanHosler p All Supervisory Enginee1* ~ L - - ')../o/Jt Materials Engineering Sal'al1 Dividsaver. R All Plinolpal Englneei* * ., . )aA.4 f}a,v; drau-'1 z/1/JB Materials Engineering' Pavan Thallapragada A All Manager MSAU 1,~tl~ 2)11,/1t

(

Note: P/LP .9.esignates Preparer (P), Lead Pr.eparer (LP)

M designates Mento1* (M) '

RILR designates Reviewer (R), Lead Iteviewe1* (LR)

A-CRF designates Project Manager Approver of Customer Requked Format (A-CRF)

A designates Appl'Ove1*/RTM-: Verification ofReviewer Independence Project Manager Approval of Customer-References (N/A if not applicable)

Name Title (printed or typed) (printed or typed) Date David Skulina Project Manager Page2 L ___ _

framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Record of Revision Revision Pages/Sections/

No. Paragraphs Changed Brief Description / Change Authorization 000 All Original submittal; Note proprietary version is 51-9271544-000.

001 Sections 1.0 and 6.0 Updated text and Figures 1-1 and 1-2 in Section 1.0, and updated References 1 and 2 to the latest revisions.

002 Multiple See Proprietary version (51-9271544-003).

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fram ome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Table of Contents Page SIGNATURE BLOCK ................................................................................................................................ 2 RECORD OF REVISION .......................................................................................................................... 3 LIST OF FIGURES ................................................................................................................................... 5 1.0 PURPOSE ..................................................................................................................................... 6 2.0 ASSUMPTIONS ............................................................................................................................ 8 2.1 Assumptions Requiring Verification ................................................................................... 8 2.2 Justified Assumptions ........................................................................................................ 8 3.0 CORROSION OF EXPOSED LOW ALLOY STEEL ..................................................................... 9 3.1 General Corrosion ............................................................................................................. 9 3.2 Galvanic Corrosion ............................................................................................................ 9 3.3 Crevice Corrosion ............................................................................................................ 10 3.4 Stress Corrosion Cracking .............................................................................................. 10 4.0 CORROSION OF ALLOY 690 A,ND ALLOY 52M ....................................................................... 12

5.0 CONCLUSION

............................................................................................................................ 13

6.0 REFERENCES

............................................................................................................................ 13 Page4

framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary List of Figures Page Figure 1-1: Original Configuration [1,2) .................................................................................................... 7 Figure 1-2: Final Repair Configuration [1,2] ............................................................................................. 8 Figure 3-1:

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fram Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-0 Reactor Vessel Nozzle Modification - Non-Proprietary 1.0 PURPOSE The repair of the N16-D reactor vessel instrumentation nozzle in the Limerick Generating Station Unit 2 (LGS-2) reactor vessel changed the penetration configuration in the following ways: 1) the repair exposed the SA-533 Grade B, Class 1 low alloy steel reactor vessel and E8018-NM low alloy steel weld pad to water conditions, 2) included a new Alloy 690 nozzle as part of the pressure boundary, and 3) included a new Alloy 52M weld pad and partial penetration J-groove weld as part of the pressure boundary [1,2]. Also, the reducing insert to nozzle weld is now an Alloy 52M dissimilar metal weld. The original configuration and the final repair configuration are shown in Figure 1-1 and Figure 1-2, respectively. Note that the weld joining the stainless steel reducing insert (i.e., Grade F316) to the stainless steel pipe in the original configuration is assumed to be stainless steel because the joined base metals are both stainless steel.

The following corrosion evaluation considers potential material degradation due to each of these changes. For each degradation mechanism, the evaluation will either justify that the degradation mechanism is not an issue for the remaining life of the plant or the evaluation will conclude that further analysis is required.

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framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Figure 1-1: Original Configuration [1,2]

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framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Figure 1-2: Final Repair Configuration [1,2]

2.0 ASSUMPTIONS 2.1 Assumptions Requiring Verification There are no assumptions requiring verification.

2.2 Justified Assumptions The weld joining the stainless steel reducing insert to the stainless steel pipe in the original configuration is assumed to be stainless steel because the joined base metals are both stainless steel.

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framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary 3.0 CORROSION OF EXPOSED LOW ALLOY STEEL The low alloy steel reactor vessel material exposed due to the repair will be in the water space environment given the elevation of the N16-D nozzle. LGS-2 implements the water chemistry control requirements ofBWRVIP-190 Revision 1 to mitigate corrosion [3, 4].

3.1 General Corrosion 3.2 Galvanic Corrosion Page 9

am Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary

[

l 3.3 Crevice Corrosion

[

] The environmental conditions in a crevice can become aggressive with time and can cause accelerated local corrosion.

[

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The test results are supported by operating experience (and simulated operating experience) in light water reactors. [

]

3.4 Stress Corrosion Cracking Although it is very unlikely that SCC cracks will initiate and propagate in low alloy steel under normal BWR conditions, it is impossible to completely rule out. Hence, it is prudent to examine the feasibility of performing an allowable flaw size evaluation by applying the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI criteria [9]. [

.]

[

]

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framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Page 11

framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Figure 3-1: [

]

4.0 CORROSION OF ALLOY 690 AND ALLOY 52M Stress corrosion cracking failures of Alloy 600 and its associated weld metals (Alloy 82/182) have occurred in domestic and international light water reactors. The BWR industry addressed this issue by replacing or modifying affected materials with a modified version of Alloy 82 [14]. The modified version of Alloy 82 adds carbide stabilizers (Niobium and Titanium) to minimize chromium depletion at the grain boundaries. The PWR industry selected Alloy 690 and Alloy 52/152 as replacement materials [15]. Alloy 690 was also thermally treated to improve the microstructure, but grain boundary chromium depletion of Alloy 690/52/152 was avoided by doubling the chromium content (from -15% to -30%) instead of using carbide stabilizers. Alloys 690/52/152 have been in service for decades with no reported failures. Laboratory studies indicate that Alloy 690 and Alloy 52/152 have superior SCC resistance relative to the Alloy 600 and Alloy 82/182.

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fram ome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification - Non-Proprietary Although most testing of Alloy 690/52/152 has been under PWR conditions, some studies have been performed in environments more similar to BWRs. Creviced U-bend specimens of Alloy 600 and Alloy 690 were tested at 600°F for 48 weeks with an environment of 6 ppm oxygen [16]. The Alloy 600 readily cracked, whereas Alloy 690 showed no cracking. Also, testing of Alloy 690 in high purity water containing 36 ppm oxygen at 289°C

(-550°F) for 47 weeks resulted in no cracking [16].

Extensive testing has been performed on Alloy 52/152 in high temperature deaerated water, which indicate that Alloy 52/152 is much less susceptible to SCC compared to Alloy 82/182 (the Alloy 600 weld metal) [15, 17, 18].

Test data of Alloy 52/152 in a high temperature oxygenated environment is not readily available, but Alloy 52/152 is expected to have a low susceptibility to SCC under these conditions as well, based on the similarity of Alloy 52/152 to Alloy 690.

The only difference between the Alloy 52M to be used in the repair and Alloy 52/152 are small alloying additions to improve weldability. The corrosion resistance is similar.

5.0 CONCLUSION

The modification of the N16-D reactor vessel nozzle at LGS-2, which exposed the low alloy steel reactor vessel to a water environment and introduced new materials (Alloy 690 and Alloy 52M), is found acceptable for the remaining life of the plant for all degradation mechanisms considered with the exceptions of general corrosion and SCC of the exposed low alloy steel, both of which require disposition by additional analyses. [

] Based on laboratory studies and operating experience, the replacement higher chromium content nickel-based alloys (Alloy 690 and Alloy 52M) have a high resistance to SCC.

6.0 REFERENCES

References identified with an (*) are maintained within Exelon Records System and are not retrievable from Framatome Records Management. These are acceptable references per Framatome Administrative Procedure 0402-01, Attachment 8. See page 2 for Project Manager Approval of customer references.

I. [

]

2. [

]

3. *BWRVIP-190 Revision 1: BWR Vessel and Internals Project, Volume 2: BWR Water Chemistry Guidelines - Technical Basis. EPRI, Palo Alto, CA: 2014. 3002002623.

4. LGS UFSAR -Appendix A, "Updated Final Safety Analysis Report Supplement," Revision 18.

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framatome Document No.: 51-9271770-002 Corrosion Evaluation of the Limerick Unit 2 N16-D Reactor Vessel Nozzle Modification

5. [ ]

6. D.C. Vreeland, et al., "Corrosion of Carbon and Low-Alloy Steels in Out-of-Pile Boiling-Water-Reactor Enviromnent," Corrosion, Vol. 17, No. 6, 1961.

7. [

]

8. [

]

9. AS.ME Boiler and Pressure Vessel Code,Section XI, 2007 Edition, including Addenda through 2008.

10. H.P. Seifert and S. Ritter, "New Observations about the sec Crack Growth Behavior of Low-Alloy RPV Steels under B WR/NWC Conditions," 11th International Conference on Enviromnental Degradation of Materials in Nuclear Power Systems, Aug 10-14, 2003, Stevenson, WA, ANS.

11. [

]

12. License Renewal Application, Limerick Generating Station Units 1 and 2, Facility Operating License Nos. NPF-39 and NPF-85, June 22, 2011, NRC AccessionNmnber ML1l 179A10I.

13. [

]

14. NUREG/CR-6923, "Expert Panel Report on Proactive Materials Degradation Assessment."

15. .Afaterials Reliability Program (MRP):Resistance to Prima,y Water Stress Corrosion Cracking ofAlloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111), EPRI, Palo Alto, CA: 2004. 1009801.

16. Sedriks, A.J., Schultz, J. W., Cordovi, M.A., "Inconel Alloy 690 - A New Con'Osion Resistant Material,"

Corrosion Engineering (Boshoku Gijutsu), vol. 28, pp. 82-95, 1979, Japan Society of Corrosion Engineering.

17.

• M. J. Psaila-Dombrowski et al., "Evaluation of Weld Metals 82, 152, 52 and Alloy 690 Stress Corrosion Cracking and Corrosion Fatigue Susceptibility," Eighth International Symposium on Environmental Degradation of Materials In Nuclear Power Systems - Water Reactors, Aug 10-14 1997, Amelia Island, FL,ANS.

18. Crum, J.R., Nagashima, T., "Review of Alloy 690 Steam Generator Studies," Eighth International Symposium on Enviromnental Degradation of Materials In Nuclear Power Systems - Water Reactors, Aug 10-14 1997, Amelia Island, FL, ANS.

The Vice President of Products and Engineering has approved the use of Reference 5.

Cruel Fisher (Vice President of Products and E n g i n e e r ~ ( o c \ psl, f' Page 14