Inhibiting stress corrosion cracking in the primary coolant circuit of a nuclear reactor

• US 5,108,697 A
• Filed: 10/19/1990
• Issued: 04/28/1992
• Est. Priority Date: 10/19/1990
• Status: Expired due to Term

• Current Assignee: Westinghouse Electric Company LLC (Brookfield Asset Management Incorporated)
• Original Assignee: Westinghouse Electric Company LLC (Brookfield Asset Management Incorporated)
• Inventors: Esposito, Judith B., Pement, Fredric W., Esposito, John N.
• Primary Examiner(s): Walsh, Donald P.
• Assistant Examiner(s): CHELLIAH, MEENA A
• Application Number:
  US07/600,465
• Time in Patent Office: 557 Days
• Field of Search: 376/306, 376/305, 376/906, 376/312, 376/310, 148/162, 148/428, 148/426, 148/427, 420/441, 420/442, 420/459, 252/631, 252/389.52, 252/400.52
• US Class Current: 376/306
• CPC Class Codes: C23F 11/18 using inorganic inhibitors
• CPC Class Codes: G21C 19/28 Arrangements for introducing fluent material into the reactor core; Arrangements for removing fluent material from the reactor core pumping coolant G21D
• CPC Class Codes: Y02E 30/30 Nuclear fission reactors

Abstract

A primary coolant circuit for cooling a nuclear reactor has wetted mechanically stressed nickel base alloy components such as Alloy 600 tubes in steam generators having oxidized surfaces comprising 1-10 w/o zinc, which tubes are inhibited against primary water stress corrosion cracking. The crack initiation times may be delayed by a factor of two in pressurized water nuclear reactors.

Claims

1. A method of inhibiting stress corrosion cracking in a pressurized water nuclear reactor, comprising the steps of:

(a) circulating coolant water through the primary cooling circuit of a pressurized water nuclear reactor in contact with nickel base alloys comprising at least about 60 w/o nickel, about 13-30 w/o chromium and about 0.3-12 w/o iron,

(b) maintaining the coolant water composition at

about 1-2500 ppm boron,

about 0.1-10 ppm lithium, and

about 15-75 cc (STP) hydrogen/kg water,

(c) maintaining the coolant water pH between about 5.2 and about 7.4 at 20°

C. and conductivity between about 1 and 30 μ

S/cm at 20°

C.; and

(d) adding zinc to the coolant water in amounts sufficient to maintain the zinc concentration between about 5 and about 1000 ppb.

2. The method of claim 1, wherein

the coolant water is maintained between about 100 and 1000 ppm boron and between about 0.1 and 3 ppm lithium and at a pH between about 6.0 and 7.2, and

the zinc is added to the coolant water in amounts sufficient to maintain a zinc concentration between about 10 ppb and about 200 ppb.

20. The method of claim 1, wherein the zinc is added as zinc borate.

3. A method of inhibiting stress corrosion cracking of nickel alloy components in a steam generator in a primary coolant circuit of a nuclear reactor, comprising the steps of:

oxidizing the surfaces of components disposed in a steam generator in a primary coolant circuit of a nuclear reactor which comprise at least about 60 w/o nickel, about 13-30 w/o chromium and about 0.3-12 w/o iron;

introducing zinc into the top 50 Angstroms of the oxidized surface to produce a zinc concentration therein of at least 1 w/o; and

circulating primary coolant through the circuit at a temperature greater than about 330°

C.

4. The method of claim 3, by producing the 1 w/o zinc concentration before circulating the primary coolant.

5. The method of claim 4, by circulating primary coolant containing less than 1 ppb zinc.

6. The method of claim 3, by producing the 1 w/o zinc concentration with the circulating primary coolant.

7. The method of claim 3, by oxidizing and introducing zinc into the surfaces of the components to produce in the top 50 Angstroms of the surfaces an oxide comprising at least about 15 w/o nickel, about 5-20 w/o chromium, about 2-20 w/o iron, about 3-7 w/o zinc and at least about 25 w/o oxygen as determined by electron spectroscopy.

8. The method of claim 3, by circulating primary coolant as an aqueous solution containing about 1-2500 ppm boron, about 0.1-10 ppm lithium, about 15-75 cc (STP) hydrogen/kg solution, and about 5-20 ppb zinc, and has a pH at 20°

C. of about 5.2-7.4 and a conductivity at 20°

C. of about 1-30 μ

S/cm.

9. A nuclear reactor having a primary coolant circuit with mechanically stressed components wetted by a primary coolant, which components are fabricated from an alloy comprising at least about 60 w/o nickel, about 13-30 w/o chromium and about 0.3-12 w/o iron and have oxidized surfaces wetted by the primary coolant comprising about 1-10 w/o zinc (as determined by electron spectroscopy) in their top 50 Angstroms, the oxidized surfaces of the components at a temperature of at least about 330°

C.

10. A steam generator in a primary coolant circuit of a nuclear reactor having mechanically stressed components wetted by a primary coolant;

which components are fabricated from an alloy comprising at least about 60 w/o nickel, about 13-30 w/o chromium and about 0.3-12 w/o iron and have oxidized surfaces wetted by the primary coolant comprising about 1-10 w/o zinc (as determined by electron spectroscopy) in their top 50 Angstroms.

11. The steam generator of claim 10, wherein the top 50 Angstroms of the oxidized surfaces of the mechanically stressed components comprise at least about 15 w/o nickel, about 5-20 w/o chromium, about 2-20 w/o iron, about 3-7 w/o zinc and at least about 25 w/o oxygen.

12. The steam generator of claim 11, wherein the top 50 Angstroms of the oxidized surfaces of the mechanically stressed components comprise about 24-36 w/o nickel, about 8-12 w/o chromium, about 5-10 w/o iron, about 4-6 w/o zinc and about 35-45 w/o oxygen.

13. The steam generator of claim 11, wherein the mechanically stressed components are fabricated from alloys comprising 14-17 w/o chromium, 6-10 w/o iron, up to 0.5 w/o silicon, up to 10 w/o manganese, no more than 0.5 w/o carbon, no more than 0.015 w/o sulfur and the balance nickel, the nickel concentration being at least 72 w/o.

14. The steam generator of claim 10, wherein the components comprise bent tubes.

15. The steam generator of claim 14, wherein the tubes have U bends.

16. The steam generator of claim 10, wherein the components comprise expanded tubes.

17. The steam generator of claim 10, wherein the components comprise tube plugs.

18. The steam generator of claim 10, in a pressurized water reactor system.

19. The steam generator of claim 10, wherein the components are at a temperature greater than about 330°

C.