물 누출에 의한 2.5Cr-1Mo 누출 Simulator 반응부위에서 나트륨-물 반응특성

정경채; 정지영; 김태준; 최종현; 황성태; Kyung Chai Jeong; Ji Young Jeong; Tae Joon Kim; Jong hyun Choi; Sung Tai Hwang

Journal of the Korean Industrial and Engineering Chemistry, Vol.12, No.7, 738-743, November, 2001

Export Citation

물 누출에 의한 2.5Cr-1Mo 누출 Simulator 반응부위에서 나트륨-물 반응특성

Sodium-Water Reaction Characteristics at the Reaction Site of a 2.25Cr-1Mo Simulator by Water Leakage

정경채^†, 정지영, 김태준, 최종현, 황성태

한국원자력연구소 KALIMER 기술개발팀, 대전 305-600

Kyung Chai Jeong^†, Ji Young Jeong, Tae Joon Kim, Jong hyun Choi, and Sung Tai Hwang

KALIMER Technology Development Team, Korea Atomic Energy Research Institute, Taejeon 305-600, Korea

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초록

2.25Cr-1Mo steel 시편을 사용해서 고온의 액체 나트륨 분위기에서 물 누출실험을 통해 나트륨-물 반응특성 및 재질의 부식 현상에 대한 연구를 수행하였다. Self-plugging 되었던 누출경로는 나트륨-물 반응에 의해 생성된 나트륨 산화물과 튜브 재질과의 반응에 의해 형성된 부식 생성물에 의해 plugging 되는 것으로 추측되었고, reopen 현상은 초기 누출량과는 무관한 것으로 해석되었으며, 최소 50 min에서 최대 380 min까지의 reopen time을 나타내었다. Seat 표면에서의 부식생성물은 튜브 재질의 모체중에 있는 Fe나 Cr의 용해물과 반응해서 (NaOH+Na2O)·FexOy 화합물이나 (NaCrO2+Na2CrO4)와 같은 복잡한 화합물이 생성되는 것으로 예측되었으며, Na은 모체의 matrix 내로는 penetration 되지 않는 것으로 판단되었다. 나트륨-물 반응에 의한 부식부위에서 Fe, Cr Ni성분의 감소현상이 확인되었는데, 정확한 감소 메커니즘은 확인할 수 없었으나, 나트륨-물 반응에 의해 발생된 고온의 반응열이 반응부위에서 발생된 나트륨 산화물과 화학반응을 일으키면서 segregation이 발생된 것으로 판단되었다.

In this study, experiments of sodium-water reaction and corrosion phenomena at the surface of water infecting simulator, 2.25Cr-1Mo steel, were carried out in liquid sodium atmosphere. The self-plugging phenomena at leak path of the simulator was caused by sodium-water reaction products and corrosion precipitates from the reaction of tube material components and sodium compounds. The reopen phenomena at plugged leak path of the simulator was observed from 50 to 380 min and it was determined to be independent of initial leak rate. Corrosion precipitates, which were produced by the reaction of sodium and dissolved components of the tube material, such as Fe and/or Cr, were expected to have complex forms of (NaOH+Na2O)·FexOy and (NaCrO2+Na2CrO4) compounds. Liquid sodium did not penetrate into the 2.25Cr-1Mo steel surface matrix. Decrease of Fe, Cr, and Ni contents, at the corrosion site of leak simulator, were analyzed, but its mechanism was not fully understood. It is believed that the segregation phenomenon was due to the exothermic heat; the corrosive sodium compounds formed by Na-H2O reaction: and the deposits produced by reaction products and material matrix.

Keywords:sodium-water reaction;wastage;water leak;corrosion

[References]

Biscarel J, Hugla M, Peturaud P, Study of the effects of major water leakage in sodium heated steam generator, International Topical Meeting on Liquid Metal Fast Breeder Reactor Safety and Operational Aspects, Conf-760503 Lyon, France (1982)
Quinet JL, Lannou L, Nucl. Technol., 46, 582 (1979)
Addison CC, The Chemistry of the Liquid Alkali Metals, John Wiley & Sons, New York (1984)
Greene DA, Nucl. Technol., 14, 218 (1972)
Judd AM, Currie R, Linekar GAB, Henderson JDC, Nucl. Energy, 31(2), 221 (1992)
Sane JO, Regimbal JJ, Fairbairn JA, Meyer RA, Analysis of the U.S. Large leak test program series I, Sodium-water reaction test results, Proceeding of the International Meeting of Fast Reactor Safety Technology, Seattle (1979)
Greene DA, Sodium-water reaction phenomena associated small leaks in LMFBR steam generator, Liquid Metal Engineering and Technology Proceeding 3rd International Conference. p. 13, Oxford (1984)
Currie R, Innes N, Fraser AD, LIMET '88, Proceedings 4th International Conference Liquid Metal Engineering Technology, Avignon, France (1988)
Avatteri C, "Sodium: Characteristics and Standards for Its Use," Harwood Academic Publishers (1981)
Nei H, J. Nucl. Sci. Technol., 14(9), 652 (1977)
Fidler RS, Saagi R, Nucl. Energy, 23(6), 385 (1984)
Schnitker W, Atomkernenergie, 24, 233 (1974)
Shimoyama K, Kuroha M, 日本原子力學會(昭 61 分科會豫告集), Paper C39 (1986)
Ford JA, Literature review of sodium-water reactions, Atomic Power Development Association, Report-167 (1965)
Newman RE, Smith CA, J. Nucl. Mater., 52, 173 (1974)
IAEA, "Fast Reactor Database," IAEA-TECDOC-866, 104 (1996)

[Referenced By]

[1] Biscarel J, Hugla M, Peturaud P, Study of the effects of major water leakage in sodium heated steam generator, International Topical Meeting on Liquid Metal Fast Breeder Reactor Safety and Operational Aspects, Conf-760503 Lyon, France (1982)

[2] Quinet JL, Lannou L, Nucl. Technol., 46, 582 (1979)

[3] Addison CC, The Chemistry of the Liquid Alkali Metals, John Wiley & Sons, New York (1984)

[4] Greene DA, Nucl. Technol., 14, 218 (1972)

[5] Judd AM, Currie R, Linekar GAB, Henderson JDC, Nucl. Energy, 31(2), 221 (1992)

[6] Sane JO, Regimbal JJ, Fairbairn JA, Meyer RA, Analysis of the U.S. Large leak test program series I, Sodium-water reaction test results, Proceeding of the International Meeting of Fast Reactor Safety Technology, Seattle (1979)

[7] Greene DA, Sodium-water reaction phenomena associated small leaks in LMFBR steam generator, Liquid Metal Engineering and Technology Proceeding 3rd International Conference. p. 13, Oxford (1984)

[8] Currie R, Innes N, Fraser AD, LIMET '88, Proceedings 4th International Conference Liquid Metal Engineering Technology, Avignon, France (1988)

[9] Avatteri C, "Sodium: Characteristics and Standards for Its Use," Harwood Academic Publishers (1981)

[10] Nei H, J. Nucl. Sci. Technol., 14(9), 652 (1977)

[11] Fidler RS, Saagi R, Nucl. Energy, 23(6), 385 (1984)

[12] Schnitker W, Atomkernenergie, 24, 233 (1974)

[13] Shimoyama K, Kuroha M, 日本原子力學會(昭 61 分科會豫告集), Paper C39 (1986)

[14] Ford JA, Literature review of sodium-water reactions, Atomic Power Development Association, Report-167 (1965)

[15] Newman RE, Smith CA, J. Nucl. Mater., 52, 173 (1974)

[16] IAEA, "Fast Reactor Database," IAEA-TECDOC-866, 104 (1996)