ML11132A184
| ML11132A184 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/11/2011 |
| From: | Xu H AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 51-9156228-000 | |
| Download: ML11132A184 (23) | |
Text
ATTACHMENT (7)
CORROSION EVALUATION FOR PRESSURIZER HEATER SLEEVE REPAIR - NON-PROPRIETARY Calvert Cliffs Nuclear Power Plant, LLC May 11, 2011
A 20004-018 (10/18/2010)
AREVA AREVA NP Inc.
Engineering Information Record Document No.: 51 - 9156228 - 000 Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair Page 1 of 22
A 20004-018 (10/18/2010)
Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair Safety Related? Z YES -- NO Does this document contain assumptions requiring verification? - YES Z NO Does this document contain Customer Required Format? YES NO Signature Block PILP, RILR, Pages/Sections Name and A/A-CRF, Prepared/Reviewed/
Title/Discipline Signature A/A-CRI Date, Approved or Comments Hongqing Xu P 3 All Principal Engineer Sarah Davidsaver. R All Materials Engineer IV- - 1 Brian Haibach 7i7,7,/ A / All Unit Manager Note: P/LP designates Preparer (P), Lead Preparer (LP)
R/LR designates Reviewer (R), Lead Reviewer (LR)
A/A-CRF designates Approver (A), Approver of Customer Requested Format (A-CRF)
A/A-CRI designates Approver (A), Approver - Confirming Reviewer Independence (A-CRI)
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A 20004-018 (10/18/2010)
Document No.: 51-9156228-000 AREVA Corrosion Evaluation for CalveDCliffs Unit 2 Pressurizer Heater Sleeve Repair Record of Revision Revision Pages/Sections/
No. Paragraphs Changed Brief Description / Change Authorization 000 All Original Release I _______ __________________
t I.
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A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair Table of Contents Page SIG NATURE BLOCK ................................................................................................................................ 2 RECO RD O F REVISIO N .......................................................................................................................... 3 LIST O F TABLES ..................................................................................................................................... 5 LIST O F FIG URES ................................................................................................................................... 6 1.0 BACKG RO UND ............................................................................................................................ 7 2 .0 S C O P E.......................................................................................................................................... 8 3.0 INDUSTRY OCCURRENCES OF EXPOSED CARBON/LOW ALLOY STEEL BASE METAL.. 12 4.0 CORROSION OF LOW ALLOY STEEL EXPOSED TO RCS ................................................ 14 4.1 General Corrosion ...................................................................................................... 14 4.1.1 General Corrosion Data ............................................................................ 14 4.1.2 Pressure Boundary Leakage (W astage) ................................................... 14 4.1.3 General Corrosion Rate for Locations A, B and C .................................... 14 4.1.4 Long Term General Corrosion ............................... 15 4.2 Crevice Corrosion ........................................................................................................ 16 4.3 Galvanic Corrosion ...................................................................................................... 16 4.4 Stress Corrosion Cracking .......................................................................................... 17 4.5 Hydrogen Em brittlem ent ............................................................................................. 17 5.0 CO RROSIO N O F ALLOYS 690/52M ..................................................................................... 18 5.1 SCC Resistance: Near Norm al Conditions ................................................................. 18 5.2 SCC Resistance: Extrem e Conditions ........................................................................ 18 5.3 Hydrogen Effects ........................................................................................................ 19 6.0 SUM MARY .................................................................................................................................. 20
7.0 REFERENCES
............................................................................................................................ 20 Page 4
A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair List of Tables Page TABLE 4-1: COMBINED AVERAGE CORROSION RATE FOR STAGNANT OR LOW FLOW C O NDITIONS ........................................................................................................................ 15 Page 5
A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair List of Figures Page FIGURE 2-1: 1990 HEATER SLEEVE REPAIR CONFIGURATION (EXCEPT H3 LOCATION) [2] ......... 9 FIGURE 2-2: 1990 REPAIR CONFIGURATION AT H3 HEATER SLEEVE LOCATION [2]................... 10 FIG U R E 2-3: P R O PO S ED N 3 R EPA IR [4]............................................................................................. 11 Page 6
A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair
1.0 BACKGROUND
Note: The AREVA NP "Proprietary" version of this document is 51-9156066-000. The AREVA NP Proprietary information is removed from this non-proprietary document as indicated by "{ }".
All 120 Alloy 600 pressurizer heater sleeves at Calvert Cliffs Nuclear Power Plant Unit 2 (CCNPP-2) were repaired in 1990. The 1990 repair configuration is described in References 1 and 2. 119 heater sleeves were replaced with Alloy 690 heater sleeves, each consisting of an outer sleeve and an inner sleeve. Since the inner sleeve is irrelevant to this evaluation, only the Alloy 690 outer sleeve is discussed in this document. The Alloy 690 outer sleeve illustrated in Figure 2-1 for the 1990 repair is hereafter referred to as "sleeve". The remaining H3 location was plugged with one oversized Alloy 690 sleeve and an Alloy 690 plug as illustrated in Figure 2-2. At each of the 120 locations, an Alloy 82 weld pad was built up on the outside surface of the bottom head. The Alloy 690 sleeve was attached to the weld pad with an Alloy 82 partial penetration weld on the outside surface of the bottom head. On the inside surface (i.e., cladding surface) of the pressurizer bottom head, an autogenous weld was used to seal the annular crevice between the Alloy 690 sleeve and penetration at each location.
In February 2011, the Alloy 82 partial penetration weld at heater sleeve N3 location was found leaking reactor coolant system (RCS) water. {
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A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair 2.0 SCOPE This document evaluates the following corrosion concerns for the 1990 heater sleeve repair and the proposed repair at the N3 heater location in 2011:
- 1. Corrosion evaluation of the low alloy steel (LAS) base metal of the pressurizer bottom head (SA-533, Grade B, Class 1)[3] exposed to the RCS water. (Sections 3 and 4)
- a. General Corrosion
- b. Crevice Corrosion
- c. Galvanic Corrosion
- e. Hydrogen Embrittlement The exposed LAS base metal locations are shown as Locations A, B, and C in Figure 2-1, Figure 2-2, and Figure 2-3.
- The exposed LAS {
}
- The exposed LAS {
}
- The exposed LAS {
}
- 2. Corrosion evaluation of Alloy 690 and Alloy 52M for the proposed N3 repair shown in Figure 2-3. (Section 5)
Note: The following are not covered by this document:
" Corrosion evaluation for the remaining Alloy 82 weld pad after the proposed N3 heater sleeve repair.
- Corrosion evaluation for the 1990 repair of heater sleeves, except for the corrosion evaluation of the exposed LAS base metal.
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair
{ }
Figure 2-1: 1990 Heater Sleeve Repair Configuration (except H3 location) [2]
(Note: only the outer sleeve is shown for the 1990 repair)
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A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair
{ }
Figure 2-2: 1990 Repair Configuration at H3 Heater Sleeve Location [2]
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair
{ }
Figure 2-3: Proposed N3 Repair [4]
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A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair 3.0 INDUSTRY OCCURRENCES OF EXPOSED CARBON/LOW ALLOY STEEL BASE METAL The carbon or LAS components in the pressurizer, reactor vessel, and the steam generator exposed to PWR RCS primary coolant are clad with either stainless steel or nickel-base alloy in order to prevent corrosion of the carbon or LAS base metal. Throughout the operating history of domestic PWRs, there have been many cases where a localized area of the carbon or LAS base metal has been exposed to the PWR RCS primary coolant due to damage to the cladding or repair configuration. Several such instances are listed below:
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair
}
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair 4.0 CORROSION OF LOW ALLOY STEEL EXPOSED TO RCS Several types of corrosion can potentially occur when carbon and LAS base metal are exposed to primary coolant. During operation, the primary coolant is deaerated and at -650°F in the pressurizer. During normal operation the primary coolant inside Locations A, B, and C will be stagnant. During shutdown, the primary coolant temperature is at 70-100°F and may become aerated. The following subsections discuss the potential corrosion mechanisms for the exposed LAS base metal at Locations A, B, and C.
4.1 General Corrosion General corrosion is defined as uniform deterioration of a surface by chemical or electrochemical reactions with the environment. Austenitic stainless steels and nickel-base alloys (e.g. wrought Type 304, Type 316, Alloy 600, and Alloy 690 materials and their equivalent weld metals) are virtually immune to general corrosion from exposure to PWR primary coolant. Carbon and LAS, however, may be subject to general corrosion upon exposure to primary coolant. The general corrosion rates of carbon and LAS during aerated and deaerated reactor coolant conditions are discussed below.
4.1.1 General Corrosion Data
{
}
4.1.2 Pressure Boundary Leakage (Wastage)
{
}
4.1.3 General Corrosion Rate for Locations A, B and C
{
}
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A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair Table 4-1: Combined Average Corrosion Rate for Stagnant or Low Flow Conditions Operating Condition Shutdown Condition Combined Average Corrosion 24 month cycle 2 month cycle Rate
{ } { } I{ }
4.1.4 Long Term General Corrosion The corrosion rate of the exposed LAS at Locations A, B and C is conservatively estimated to {
} for a 24-month fuel cycle (24-month operation followed by a 2-month shutdown) as described above.
The Fe release rate into the RCS water from these locations is evaluated below.
Location A
{
I Location B
{
}
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair Location C
{
}
4.2 Crevice Corrosion
{
}
4.3 Galvanic Corrosion
{
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair
}
4.4 Stress Corrosion Cracking Stress corrosion cracking (SCC) can occur only when the following three conditions are present: (1) a susceptible material, (2) a tensile stress, (3) and an aggressive environment.
{
Ebrittemen Hydroen 4.5 Hydrogen Emnbrittlemnent
{
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair 5.0 CORROSION OF ALLOYS 690/52M The corrosion resistance of Alloy 690 and weld metal Alloys 52 and 152 has been extensively studied as a result of numerous PWSCC failures in mill annealed Alloy 600 and weld metals Alloys 82 and 182 in primary water environments. As a result of Alloy 600 and Alloys 82 and 182 failures, Alloy 690 and Alloys 52 and 152 have been chosen by the nuclear power industry as the replacement material of choice for Alloy 600 and Alloy 82 and 182 components and welds in PWRs. Alloy 52M, a modified version of Alloy 52 for improved weldability, is the same as Alloy 52 in terms of the corrosion resistance properties discussed below.
5.1 SCC Resistance: Near Normal Conditions
{
}
5.2 SCC Resistance: Extreme Conditions
{
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A Document No.: 51-9156228-000 AREVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair
}
5.3 Hydrogen Effects f
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A Document No.: 51-9156228-000 AR EVA Corrosion Evaluation for Calvert Cliffs Unit 2 Pressurizer Heater Sleeve Repair 6.0
SUMMARY
The preceding sections have evaluated the potential corrosion mechanisms affecting the exposed LAS in the 1990 heater sleeve repair and the proposed N3 pressurizer heater sleeve repair in 2011 for the CCNPP-2 pressurizer.
For the LAS exposed to the primary coolant, galvanic corrosion, hydrogen embrittlement, SCC, and crevice corrosion are not expected to be a concern. Based on industry data and AREVA NP's experience, the general corrosion rate of exposed LAS at Locations A, B and C is conservatively estimated to { }
for a 24-month fuel cycle (24-month operation followed by a 2-month shutdown). Due to the {
} at Location A is expected to be quickly { ). The exposed area at Locations B and C are very small with the combined estimated Fe release rate of {
Wrought Alloy 690 has been shown to exhibit superior PWSCC resistance by extensive laboratory testing and in-reactor operating experience. There have been no reports of in-service failures of {
}. No cracking of Alloy 690 has been observed, even for oxygenated conditions {
}. However, cracking of Alloy 690 is possible under more extreme environmental conditions not found in PWRs. Alloys 52 and 152 have also been shown to exhibit superior PWSCC resistance compared to Alloys 82 and 182, particularly { }. The Alloy 690 and its weld metal Alloy 52M used for the proposed N3 repair are not likely to degrade from exposure to the primary coolant during the remainder of the service life at CCNPP-2.
7.0 REFERENCES
- 1. AREVA NP Document 08-1176078 (both Revisions 000 and 001), "Repair/Replacement of Pressurizer Heater Sleeves."
- 2. AREVA NP Document 02-1196380E-002, "Heater Sleeve Assembly & Welding Specification Drawing."
- 3. AREVA NP Document 08-9155575-000, "Calvert Cliffs Unit 2 Pressurizer Penetration N3 Modification."
- 4. AREVA NP Document 02-9155725E-000, "Calvert Cliffs Unit 2 Pressurizer Penetration N3 Plug."
- 5. Whitman, G.D., et al., A Review of Current Practice in Design, Analysis, Materials, Fabrication, Inspection, and Test, ORNL-NSIC-21, ORNL, December 1967.
- 6. Vreeland, D.C., et al., "Corrosion of Carbon and Low-Alloy Steels in Out-of-Pile Boiling Water Reactor Environment," Corrosion, v17(6), June 1961, p. 269.
- 7. Vreeland, D.C., et al., "Corrosion of Carbon and Other Steels in Simulated Boiling Water Environment:
Phase I1," Corrosion, v19(10), October 1962, p. 368.
- 8. Uhlig, H.H., and Revie, R.W., Corrosion and Corrosion Control, John Wiley & Sons, New York, 1985.
- 9. Copson, H.R., "Effects of Velocity on Corrosion by Water," Ind. Eng. Chem., v44, p. 1745, 1952.
- 10. Vreeland, D.C., "Corrosion of Carbon Steel and Low Alloy Steels in Primary Systems of Water-Cooled Nuclear Reactors," Presented at Netherlands-Norwegian Reactor School, Kjeller, Norway, August 1963.
- 11. Pearl, W.C., and G.P. Wozadlo, "Corrosion of Carbon Steel in Simulated Boiler Water and Superheated Reactor Environments," Corrosion, v21(8), August 1965, p. 260.
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- 12. Tackett, D.E., et al., "Review of Carbon Steel Corrosion Data in High Temperature, High Purity Water in Dynamic Systems," USAEC Report, WAPD-LSR (C)-134, Westinghouse Electric Corporation, October 14,1955.
- 13. DePaul, E.J., ed., "Corrosion and Wear Handbook for Water Cooled Reactors," USAEC Report, TID-7006, 1957.
- 14. Ruther, W.E., and Hart, R.K., "Influence of Oxygen on High Temperature Aqueous Corrosion of Iron,"
Corrosion, v19(4), April 1963, p. 127t.
- 15. Howells, E., and Vaughan, L.H., "Corrosion of Reactor Materials in Boric Acid Solutions," RDE-1086, Babcock &Wilcox Company, Alliance, Ohio, August 1960.
- 16. "Boric Acid Corrosion Guidebook, Revision 1,"Managing Boric Acid Corrosion Issues at PWR Power Stations, EPRI, Palo Alto, CA: 2001. 1000975.
- 17. "Evaluation of Yankee Vessel Cladding Penetrations," Yankee Atomic Electric Company to the U.S.
Atomic Energy Commission, WCAP-2855, License No. DPR-3, Docket No. 50-29, October 15, 1965.
- 18. "Absorption of Corrosion Hydrogen by A302B Steel at 70F to 500F," WCAP-7099, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1967.
- 19. AREVA NP Document 02-9156189B-000, "Weld Prep Sketch."
- 20. AREVA NP Document 02-1196376E-001, "Final Bore and Weld Prep Machining."
- 21. Ferguson, M., "Examination of 24-inch Tube Sheet Assembly from the 37-Tube OTSG", LR:68:2218-05:1, Babcock & Wilcox, AREVA NP, Inc. Proprietary, Alliance, Ohio, January 18, 1968.
- 23. P.M. Scott and D.R. Tice, "Stress Corrosion in Low Alloy Steels", Nuclear Engineering and Design, Volume 119, 1990.
- 24. J.F. Hall, "Low-Alloy Steel Component Corrosion Analyses Supporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement Programs", CE NPSD-1198-NP, Revision 00, February 15, 2001, Non-Proprietary version available for viewing on the NRC website (ML010540212).
- 25. Information Notice No. 80-38, "Cracking in Charging Pump Casing Cladding," Nuclear Regulatory Commission, October 31, 1980.
- 26. Materials Reliability Program (MRP):Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111). EPRI, Palo Alto, CA: 2004. 1009801.
- 27. Crum, J.R., Nagashima, T., "Review of Alloy 690 Steam Generator Studies," Eighth International Symposium on Environmental Degradation of Materials In Nuclear Power Systems - Water Reactors, Aug 10-14 1997, Amelia Island, FL, ANS.
- 28. Saito, N., "Crevice Corrosion of Austenitic Alloys in High-Temperature Water," Corrosion, V54(9),
September, 1998, p. 700.
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- 29. 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.
- 30. Sedriks, A.J., Schultz, J.W., Cordovi, M.A., Inconel Alloy 690 -A New Corrosion Resistant Material, Boshoku Gijutsu, Japan Society of Corrosion Engineering, V28(2), 1979.
- 31. Yonezawa, T, et. al., "Evaluation of the Corrosion Resistance of Alloy 690," Proceedings: Workshop on Thermally Treated Alloy 690 Tubes for Nuclear Steam Generators, EPRI, NP-4665S-SR, July 1986.
- 32. Symons, D.M., "The Effect of Carbide Precipitation of the Hydrogen-Enhanced Fracture Behavior of Alloy 690," Metallurgical and Materials Transactions A, V29A, April 1998, p.1265.
- 33. Brown, C.M., Mills, W.J., "Effect of Water on Mechanical Properties and Stress Corrosion Behavior of Alloy 600, Alloy 690, EN82H Welds, and EN52 Welds," Corrosion, V55(2), February 1999.
- 34. Brown, C.M., and Mills, W.J., "Fracture Toughness, Tensile and Stress Corrosion Cracking Properties of Alloy 600, Alloy 690, and Their Welds in Water," Paper 90, Corrosion96, NACE, 1996.
- 35. Mills, W.J., and C.M. Brown, "Fracture Behavior of Nickel-based Alloys in Water," Ninth International Conference on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, August 1-5, 1999, Newport Beach, CA, TMS/NACE.
- 36. Mills, W.J. et al., "Effect of Microstructure on Low Temperature Cracking Behavior of EN82H Welds,"
Tenth International Conference on Environmental Degradation of Materials in Nuclear Power Systems--
Water Reactors, August, 2001, ANS.
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