A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors
The radiolysis of water is a significant cause of corrosion damage in the primary heat transport systems (PHTSs) of water-cooled, fission nuclear power reactors (BWRs, PWRs, and CANDUs) and is projected to be a significant factor in the evolution of corrosion damage in future fusion reactors (e.g.,...
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| Format: | Article |
| Language: | English |
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MDPI AG
2022-11-01
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| Series: | Corrosion and Materials Degradation |
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| Online Access: | https://www.mdpi.com/2624-5558/3/4/38 |
| _version_ | 1797460808905523200 |
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| author | Digby D. Macdonald George R. Engelhardt |
| author_facet | Digby D. Macdonald George R. Engelhardt |
| author_sort | Digby D. Macdonald |
| collection | DOAJ |
| description | The radiolysis of water is a significant cause of corrosion damage in the primary heat transport systems (PHTSs) of water-cooled, fission nuclear power reactors (BWRs, PWRs, and CANDUs) and is projected to be a significant factor in the evolution of corrosion damage in future fusion reactors (e.g., the ITER that is currently under development). In Part I of this two-part series, we reviewed the proposed mechanisms for the radiolysis of water and demonstrate that radiolysis leads to the formation of a myriad of oxidizing and reducing species. In this Part II, we review the role that the radiolysis species play in establishing the electrochemical corrosion potential (ECP) and the development of corrosion damage due to intergranular stress corrosion cracking (IGSCC) in reactor PHTSs. We demonstrate, that the radiolytic oxidizing radiolysis products, such as O<sub>2</sub>, H<sub>2</sub>O<sub>2</sub>, HO<sub>2</sub><sup>−</sup>, and OH, when in molar excess over reducing species (H<sub>2</sub>, H, and O<sub>2</sub><sup>2−</sup>), some of which (H<sub>2</sub>) are preferentially stripped from the coolant upon boiling in a BWR PHTS, for example, renders the coolant in many BWRs oxidizing, thereby shifting the ECP in the positive direction to a value that is more positive than the critical potential (<i>E<sub>crit</sub></i> = −0.23 V<sub>she</sub> at 288 °C) for IGSCC in sensitized austenitic stainless steel (e.g., Type 304 SS). This has led to many IGSCC incidents in operating BWRs over the past five decades that has exacted a great cost on the plant operators and electricity consumers, alike. In the case of PWRs, the primary circuits are pressurized with hydrogen to give a hydrogen concentration of 10 to 50 cm<sup>3</sup>/kgH<sub>2</sub>O (0.89 to 4.46 ppm), such that no sustained boiling occurs, and the hydrogen suppresses the radiolysis of water, thereby inhibiting the formation of oxidizing radiolysis products from water. Thus, the ECP is dominated by the hydrogen electrode reaction (HER), although important deviations from the HER equilibrium potential may occur, particularly at low [H<sub>2</sub>]. In any event, the ECP is displaced to approximately −0.85 V<sub>she</sub>, which is below the critical potential for IGSCC in sensitized stainless steels but is also more negative than the critical potential for the hydrogen-induced cracking (HIC) of mill-annealed Alloy 600. This has led to extensive cracking of steam generator tubing and other components (e.g., control rod drive tubes, pressurizer components) in PWRs that has also exacted a high cost on operators and power consumers. Although the ITER has yet to operate, the proposed chemistry protocol for the coolant places it close to a BWR operating on Normal Water Chemistry (NWC) without boiling or, if hydrogen is added to the IBED-PHTS, close to a BWR on Hydrogen Water Chemistry (HWC). In the current ITER technology, the concentration of H<sub>2</sub> in the IBED-PHTS is specified to be 80 ppb, which is the concentration that will be experienced in both the Plasma Flux Area (PFA) and in the Out of Plasma Flux Area (OPFA). That corresponds to 0.90 cc(STP) H<sub>2</sub>/KgH<sub>2</sub>O, compared with 20–50 cc(STP) H<sub>2</sub>/KgH<sub>2</sub>O employed in a PWR primary coolant circuit and 5.5 to 22 cc(STP) H<sub>2</sub>/KgH<sub>2</sub>O in a BWR on hydrogen water chemistry (HWC). We predict that a hydrogen concentration of 80 ppb is sufficient to reduce the ECP in the OPFA to a level (−0.324 V<sub>she</sub>) that is sufficient to suppress the crack growth rate (CGR) below the practical, maximum level of 10<sup>−9</sup> cm/s (0.315 mm/a) at which SCC is considered not to be a problem in a coolant circuit but, in the PFA, the ECP is predicted to be 0.380 V<sub>she</sub>, which gives a calculated standard CGR of 2.7 × 10<sup>−6</sup> cm/s. This is more than three orders in magnitude greater that the desired maximum value of 10<sup>−9</sup> cm/s. We recommend that the HWC issue in ITER be revisited to develop a protocol that is effective in suppressing both the ECP and the CGR in the PFA to levels that permit the operation of the IBED-PHTS in accordance with the experience gained in fission reactor technology. |
| first_indexed | 2024-03-09T17:10:20Z |
| format | Article |
| id | doaj.art-72a68502f046443699faf85aadc2f089 |
| institution | Directory Open Access Journal |
| issn | 2624-5558 |
| language | English |
| last_indexed | 2024-03-09T17:10:20Z |
| publishDate | 2022-11-01 |
| publisher | MDPI AG |
| record_format | Article |
| series | Corrosion and Materials Degradation |
| spelling | doaj.art-72a68502f046443699faf85aadc2f0892023-11-24T14:05:56ZengMDPI AGCorrosion and Materials Degradation2624-55582022-11-013469475810.3390/cmd3040038A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating ReactorsDigby D. Macdonald0George R. Engelhardt1Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720, USAOLI Systems, Inc., 2 Gatehall Drive, Parsippany, NJ 07054, USAThe radiolysis of water is a significant cause of corrosion damage in the primary heat transport systems (PHTSs) of water-cooled, fission nuclear power reactors (BWRs, PWRs, and CANDUs) and is projected to be a significant factor in the evolution of corrosion damage in future fusion reactors (e.g., the ITER that is currently under development). In Part I of this two-part series, we reviewed the proposed mechanisms for the radiolysis of water and demonstrate that radiolysis leads to the formation of a myriad of oxidizing and reducing species. In this Part II, we review the role that the radiolysis species play in establishing the electrochemical corrosion potential (ECP) and the development of corrosion damage due to intergranular stress corrosion cracking (IGSCC) in reactor PHTSs. We demonstrate, that the radiolytic oxidizing radiolysis products, such as O<sub>2</sub>, H<sub>2</sub>O<sub>2</sub>, HO<sub>2</sub><sup>−</sup>, and OH, when in molar excess over reducing species (H<sub>2</sub>, H, and O<sub>2</sub><sup>2−</sup>), some of which (H<sub>2</sub>) are preferentially stripped from the coolant upon boiling in a BWR PHTS, for example, renders the coolant in many BWRs oxidizing, thereby shifting the ECP in the positive direction to a value that is more positive than the critical potential (<i>E<sub>crit</sub></i> = −0.23 V<sub>she</sub> at 288 °C) for IGSCC in sensitized austenitic stainless steel (e.g., Type 304 SS). This has led to many IGSCC incidents in operating BWRs over the past five decades that has exacted a great cost on the plant operators and electricity consumers, alike. In the case of PWRs, the primary circuits are pressurized with hydrogen to give a hydrogen concentration of 10 to 50 cm<sup>3</sup>/kgH<sub>2</sub>O (0.89 to 4.46 ppm), such that no sustained boiling occurs, and the hydrogen suppresses the radiolysis of water, thereby inhibiting the formation of oxidizing radiolysis products from water. Thus, the ECP is dominated by the hydrogen electrode reaction (HER), although important deviations from the HER equilibrium potential may occur, particularly at low [H<sub>2</sub>]. In any event, the ECP is displaced to approximately −0.85 V<sub>she</sub>, which is below the critical potential for IGSCC in sensitized stainless steels but is also more negative than the critical potential for the hydrogen-induced cracking (HIC) of mill-annealed Alloy 600. This has led to extensive cracking of steam generator tubing and other components (e.g., control rod drive tubes, pressurizer components) in PWRs that has also exacted a high cost on operators and power consumers. Although the ITER has yet to operate, the proposed chemistry protocol for the coolant places it close to a BWR operating on Normal Water Chemistry (NWC) without boiling or, if hydrogen is added to the IBED-PHTS, close to a BWR on Hydrogen Water Chemistry (HWC). In the current ITER technology, the concentration of H<sub>2</sub> in the IBED-PHTS is specified to be 80 ppb, which is the concentration that will be experienced in both the Plasma Flux Area (PFA) and in the Out of Plasma Flux Area (OPFA). That corresponds to 0.90 cc(STP) H<sub>2</sub>/KgH<sub>2</sub>O, compared with 20–50 cc(STP) H<sub>2</sub>/KgH<sub>2</sub>O employed in a PWR primary coolant circuit and 5.5 to 22 cc(STP) H<sub>2</sub>/KgH<sub>2</sub>O in a BWR on hydrogen water chemistry (HWC). We predict that a hydrogen concentration of 80 ppb is sufficient to reduce the ECP in the OPFA to a level (−0.324 V<sub>she</sub>) that is sufficient to suppress the crack growth rate (CGR) below the practical, maximum level of 10<sup>−9</sup> cm/s (0.315 mm/a) at which SCC is considered not to be a problem in a coolant circuit but, in the PFA, the ECP is predicted to be 0.380 V<sub>she</sub>, which gives a calculated standard CGR of 2.7 × 10<sup>−6</sup> cm/s. This is more than three orders in magnitude greater that the desired maximum value of 10<sup>−9</sup> cm/s. We recommend that the HWC issue in ITER be revisited to develop a protocol that is effective in suppressing both the ECP and the CGR in the PFA to levels that permit the operation of the IBED-PHTS in accordance with the experience gained in fission reactor technology.https://www.mdpi.com/2624-5558/3/4/38nuclear reactorswater-cooledcorrosion potentialcrack growth ratestress corrosion cracking |
| spellingShingle | Digby D. Macdonald George R. Engelhardt A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors Corrosion and Materials Degradation nuclear reactors water-cooled corrosion potential crack growth rate stress corrosion cracking |
| title | A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors |
| title_full | A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors |
| title_fullStr | A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors |
| title_full_unstemmed | A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors |
| title_short | A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors |
| title_sort | critical review of radiolysis issues in water cooled fission and fusion reactors part ii prediction of corrosion damage in operating reactors |
| topic | nuclear reactors water-cooled corrosion potential crack growth rate stress corrosion cracking |
| url | https://www.mdpi.com/2624-5558/3/4/38 |
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