1801 R1 XI.M2: Difference between revisions
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'''XI.M2 WATER CHEMISTRY''' | '''XI.M2 WATER CHEMISTRY''' | ||
Latest revision as of 19:25, 4 October 2024
XI.M2 WATER CHEMISTRY
Program Description
The main objective of this program is to mitigate damage caused by corrosion and stress corrosion cracking (SCC). The water chemistry program for boiling water reactors (BWRs) relies on monitoring and control of reactor water chemistry based on industry guidelines such as the boiling water reactor vessel and internals project (BWRVIP)-29 (Electric Power Research Institute [EPRI] TR-103515)(Archived) or later revisions. The BWRVIP-29(Archived) has three sets of guidelines: one for primary water, one for condensate and feedwater, and one for control rod drive (CRD) mechanism cooling water. The water chemistry program for pressurized water reactors (PWRs) relies on monitoring and control of reactor water chemistry based on industry guidelines for primary water and secondary water chemistry such as EPRI TR-105714, Rev. 3(Archived) and TR-102134, Rev. 3(Archived) or later revisions.
The water chemistry programs are generally effective in removing impurities from intermediate and high flow areas. The Generic Aging Lessons Learned (GALL) report identifies those circumstances in which the water chemistry program is to be augmented to manage the effects of aging for license renewal. For example, the water chemistry program may not be effective in low flow or stagnant flow areas. Accordingly, in certain cases as identified in the GALL Report, verification of the effectiveness of the chemistry control program is undertaken to ensure that significant degradation is not occurring and the component’s intended function will be maintained during the extended period of operation. As discussed in the GALL Report for these specific cases, an acceptable verification program is a one-time inspection of selected components at susceptible locations in the system.
Evaluation and Technical Basis
- 1. Scope of Program: The program includes periodic monitoring and control of known detrimental contaminants such as chlorides, fluorides (PWRs only), dissolved oxygen, and sulfate concentrations below the levels known to result in loss of material or cracking. Water chemistry control is in accordance with industry guidelines such as BWRVIP-29 (EPRI TR-103515)(Archived) for water chemistry in BWRs, EPRI TR-105714(Archived) for primary water chemistry in PWRs, and EPRI TR-102134(Archived) for secondary water chemistry in PWRs.
- 2. Preventive Actions: The program includes specifications for chemical species, sampling and analysis frequencies, and corrective actions for control of reactor water chemistry. System water chemistry is controlled to minimize contaminant concentration and mitigate loss of material due to general, crevice and pitting corrosion and cracking caused by SCC. For BWRs, maintaining high water purity reduces susceptibility to SCC.
- 3. Parameters Monitored/Inspected: The concentration of corrosive impurities listed in the EPRI guidelines discussed above, which include chlorides, fluorides (PWRs only), sulfates, dissolved oxygen, and hydrogen peroxide, are monitored to mitigate degradation of structural materials. Water quality (pH and conductivity) is also maintained in accordance with the guidance. Chemical species and water quality are monitored by inprocess methods or through sampling. The chemical integrity of the samples is maintained and verified to ensure that the method of sampling and storage will not cause a change in the concentration of the chemical species in the samples.
- BWR Water Chemistry: The guidelines in BWRVIP-29 (EPRI TR-103515)(Archived) for BWR reactor water recommend that the concentration of chlorides, sulfates, and dissolved oxygen are monitored and kept below the recommended levels to mitigate corrosion. The two impurities, chlorides and sulfates, determine the coolant conductivity; dissolved oxygen, hydrogen peroxide, and hydrogen determine electrochemical potential (ECP). The EPRI guidelines recommend that the coolant conductivity and ECP are also monitored and kept below the recommended levels to mitigate SCC and corrosion in BWR plants. The EPRI guidelines in BWRVIP-29 (TR-103515)(Archived) for BWR feedwater, condensate, and control rod drive water recommend that conductivity, dissolved oxygen level, and concentrations of iron and copper (feedwater only) are monitored and kept below the recommended levels to mitigate SCC. The EPRI guidelines in BWRVIP-29 (TR-103515)(Archived) also include recommendations for controlling water chemistry in auxiliary systems: torus/pressure suppression chamber, condensate storage tank, and spent fuel pool.
- PWR Primary Water Chemistry: The EPRI guidelines (EPRI TR-105714)(Archived), for PWR primary water chemistry recommend that the concentration of chlorides, fluorides, sulfates, lithium, and dissolved oxygen and hydrogen are monitored and kept below the recommended levels to mitigate SCC of austenitic stainless steel, Alloy 600, and Alloy 690 components. TR-105714(Archived) provides guidelines for chemistry control in PWR auxiliary systems such as the boric acid storage tank, refueling water storage tank, spent fuel pool, letdown purification systems, and volume control tank.
- PWR Secondary Water Chemistry: The EPRI guidelines (EPRI TR-102134)(Archived), for PWR secondary water chemistry recommend monitoring and control of chemistry parameters (e.g., pH level, cation conductivity, sodium, chloride, sulfate, lead, dissolved oxygen, iron, copper, and hydrazine) to mitigate steam generator tube degradation caused by denting, intergranular attack (IGA), outer diameter stress corrosion cracking (ODSCC), or crevice and pitting corrosion. The monitoring and control of these parameters, especially the pH level, also mitigates general (for steel components), crevice, and pitting corrosion of the steam generator shell and the balance of plant materials of construction (e.g., steel, stainless steel, and copper).
- 4. Detection of Aging Effects: This is a mitigation program and does not provide for detection of any aging effects.
- In certain cases as identified in the GALL Report, inspection of select components is to be undertaken to verify the effectiveness of the chemistry control program and to ensure that significant degradation is not occurring and the component intended function will be maintained during the extended period of operation.
- 5. Monitoring and Trending: The frequency of sampling water chemistry varies (e.g., continuous, daily, weekly, or as needed) based on plant operating conditions and the EPRI water chemistry guidelines. Whenever corrective actions are taken to address an abnormal chemistry condition, increased sampling is utilized to verify the effectiveness of these actions.
- 6. Acceptance Criteria: Maximum levels for various contaminants are maintained below the system specific limits as indicated by the limits specified in the corresponding EPRI water chemistry guidelines. Any evidence of aging effects or unacceptable water chemistry results is evaluated, the root cause identified, and the condition corrected.
- 7. Corrective Actions: When measured water chemistry parameters are outside the specified range, corrective actions are taken to bring the parameter back within the acceptable range and within the time period specified in the EPRI water chemistry guidelines. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable to address the corrective actions.
- 8. Confirmation Process: Following corrective actions, additional samples are taken and analyzed to verify that the corrective actions were effective in returning the concentrations of contaminants such as chlorides, fluorides, sulfates, dissolved oxygen, and hydrogen peroxide to within the acceptable ranges. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable to address the confirmation process.
- 9. Administrative Controls: Site quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable to address administrative controls.
- 10. Operating Experience: The EPRI guideline documents have been developed based on plant experience and have been shown to be effective over time with their widespread use. The specific examples of operating experience are as follows:
- BWR: Intergranular stress corrosion cracking (IGSCC) has occurred in small- and large-diameter BWR piping made of austenitic stainless steels and nickel-base alloys. Significant cracking has occurred in recirculation, core spray, residual heat removal (RHR) systems, and reactor water cleanup (RWCU) system piping welds. IGSCC has also occurred in a number of vessel internal components, including core shroud, access hole cover, top guide, and core spray spargers (US Nuclear Regulatory Commission [NRC] Information Bulletin 80-13, US NRC Information Notice [IN 95-17], US NRC Generic Letter [GL 94-03], and NUREG-1544). No occurrence of SCC in piping and other components in standby liquid control systems exposed to sodium pentaborate solution has ever been reported (NUREG/CR-6001).
- PWR Primary System: The primary pressure boundary piping of PWRs has generally not been found to be affected by SCC because of low dissolved oxygen levels and control of primary water chemistry. However, the potential for SCC exists due to inadvertent introduction of contaminants into the primary coolant system from unacceptable levels of contaminants in the boric acid, introduction through the free surface of the spent fuel pool (which can be a natural collector of airborne contaminants), or introduction of oxygen during cooldown (US NRC IN 84-18). Ingress of demineralizer resins into the primary system has caused IGSCC of Alloy 600 vessel head penetrations (US NRC IN 96-11, US NRC GL 97-01). Inadvertent introduction of sodium thiosulfate into the primary system has caused IGSCC of steam generator tubes. The SCC has occurred in safety injection lines (US NRC INs 97-19 and 84-18), charging pump casing cladding (US NRC INs 80-38 and 94-63), instrument nozzles in safety injection tanks (US NRC IN 91-05), and safety-related SS piping systems that contain oxygenated, stagnant, or essentially stagnant borated coolant (US NRC INs 97-19). Steam generator tubes and plugs and Alloy 600 penetrations have experienced primary water stress corrosion cracking (PWSCC) (US NRC INs 89-33, 94-87, 97-88, 90-10, and 96-11; US NRC Bulletin 89-01 and its two supplements).
- PWR Secondary System: Steam generator tubes have experienced ODSCC, IGA, wastage, and pitting (US NRC IN 97-88, US NRC GL 95-05). Carbon steel support plates in steam generators have experienced general corrosion. The steam generator shell has experienced pitting and stress corrosion cracking (US NRC INs 82-37, 85-65, and 90-04).
- Such operating experience has provided feedback to revisions of the EPRI water chemistry guideline documents.
References
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, Office of the Federal Register, National Archives and Records Administration, 2005.
BWRVIP-29 (EPRI TR-103515), BWR Water Chemistry Guidelines-1993 Revision, Normal and Hydrogen Water Chemistry,(Archived) Electric Power Research Institute, Palo Alto, CA, February 1994.
BWRVIP-79, BWR Water Chemistry Guidelines,(Archived) Electric Power Research Institute, Palo Alto, CA, March 2000.
BWRVIP-130, BWR Water Chemistry Guidelines,(Archived) Electric Power Research Institute, Palo Alto, CA, October 2000.
EPRI TR-102134, PWR Secondary Water Chemistry Guideline-Revision 3,(Archived) Electric Power Research Institute, Palo Alto, CA, May 1993.
EPRI TR-105714, PWR Primary Water Chemistry Guidelines-Revision 3,(Archived) Electric Power Research Institute, Palo Alto, CA, Nov. 1995.
EPRI TR-1002884, PWR Primary Water Chemistry Guidelines,(Archived) Electric Power Research Institute, Palo Alto, CA, October 2003.
US NRC IE Bulletin 80-13, Cracking in Core Spray Spargers, U.S. Nuclear Regulatory Commission, May 12, 1980.
US NRC IE Bulletin 89-01, Failure of Westinghouse Steam Generator Tube Mechanical Plugs, U.S. Nuclear Regulatory Commission, May 15, 1989.
US NRC IE Bulletin 89-01, Supplement 1, Failure of Westinghouse Steam Generator Tube Mechanical Plugs, U.S. Nuclear Regulatory Commission, November 14, 1989.
US NRC IE Bulletin 89-01, Supplement 2, Failure of Westinghouse Steam Generator Tube Mechanical Plugs, U.S. Nuclear Regulatory Commission, June 28, 1991.
US NRC Generic Letter 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors, U.S. Nuclear Regulatory Commission, July 25, 1994.
US NRC Generic Letter 95-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, U.S. Nuclear Regulatory Commission, August 3, 1995.
US NRC Generic Letter 97-01, Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations, U.S. Nuclear Regulatory Commission, April 1,1997.
US NRC Information Notice 80-38, Cracking in Charging Pump Casing Cladding, U.S. Nuclear Regulatory Commission, October 31, 1980.
US NRC Information Notice 82-37, Cracking in the Upper Shell to Transition Cone Girth Weld of a Steam Generator at an Operating PWR, U.S. Nuclear Regulatory Commission, September 16, 1982.
US NRC Information Notice 84-18, Stress Corrosion Cracking in PWR Systems, U.S. Nuclear Regulatory Commission, March 7, 1984.
US NRC Information Notice 85-65, Crack Growth in Steam Generator Girth Welds, U.S. Nuclear Regulatory Commission, July 31, 1985.
US NRC Information Notice 89-33, Potential Failure of Westinghouse Steam Generator Tube Mechanical Plugs, U.S. Nuclear Regulatory Commission, March 23, 1989.
US NRC Information Notice 90-04, Cracking of the Upper Shell-to-Transition Cone Girth Welds in Steam Generators, U.S. Nuclear Regulatory Commission, January 26, 1990.
US NRC Information Notice 90-10, Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600, U.S. Nuclear Regulatory Commission, February 23, 1990.
US NRC Information Notice 91-05, Intergranular Stress Corrosion Cracking in Pressurized Water Reactor Safety Injection Accumulator Nozzles, U.S. Nuclear Regulatory Commission, January 30, 1991.
US NRC Information Notice 94-63, Boric Acid Corrosion of Charging Pump Casing Caused by Cladding Cracks, U.S. Nuclear Regulatory Commission, August 30, 1994.
US NRC Information Notice 94-87, Unanticipated Crack in a Particular Heat of Alloy 600 Used for Westinghouse Mechanical Plugs for Steam Generator Tubes, U.S. Nuclear Regulatory Commission, December 22, 1994.
US NRC Information Notice 95-17, Reactor Vessel Top Guide and Core Plate Cracking, U.S. Nuclear Regulatory Commission, March 10, 1995.
US NRC Information Notice 96-11, Ingress of Demineralizer Resins Increase Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations, U.S. Nuclear Regulatory Commission, February 14, 1996.
US NRC Information Notice 97-19, Safety Injection System Weld Flaw at Sequoyah Nuclear Power Plant, Unit 2, U.S. Nuclear Regulatory Commission, April 18, 1997.
US NRC Information Notice 97-88, Experiences During Recent Steam Generator Inspections, U.S. Nuclear Regulatory Commission, December 16, 1997.
NUREG-1544, Status Report. Intergranular Stress Corrosion Cracking of BWR Core Shrouds and Other Internal Components, U.S. Nuclear Regulatory Commission, March 1996.
NUREG/CR-6001, Aging Assessment of BWR Standby Liquid Control Systems, G. D. Buckley, R. D. Orton, A. B. Johnson Jr., and L. L. Larson, 1992.