Potassium Hydroxide for PWR Primary Coolant ph T Control Feasibility Assessment

Transcription

1 Potassium Hydroxide for PWR Primary Coolant ph T Control Feasibility Assessment Keith Fruzzetti Technical Executive International Light Water Reactor Materials Reliability Conference and Exhibition 2016 August 1-4, 2016 Chicago, IL Co-authors: Chuck Marks and Jeff Reinders Dominion Engineering, Inc. Joel McElrath, Daniel M. Wells, Paul Frattini, Al Ahluwalia, Ryan Wolfe EPRI

2 PWRs Currently Need Highly Enriched Li-7 ph T required to be 7.0 during cycle operation Maintained by proper Li concentration Natural lithium is mostly Li-7 and Li-6 But Li-6 neutron activation generates tritium 3 6 Li + n 2 4 He H Need enriched Li-7 (99.99%) Li-7 is generated in coolant via neutron reaction with B B + n 7 3 Li + α Natural Lithium Isotope Abundance (at%) Li Li

3 Vulnerability of Li-7 Supply Realized United States Government Accountability Office 2013 report identified the concern that the required Li-7 may at some point be in short supply 1 In a new report the Government Accountability Office (GAO) raises serious concerns about the future U.S. supply of Lithium-7, a critical radioactive isotope required for the safe operation of more than half of the nation s nuclear power plants. 2 In 2015, some plants (U.S. and non-u.s.) reported an inability to procure Li-7 Still feeling the effect as full supply is being re-established The US Department of Energy (DOE) has been preparing an emergency reserve EPRI has initiated a number of activities to address this vulnerability 1 GAO , Managing Critical Isotopes: Stewardship of Lithium-7 Is Needed to Ensure a Stable Supply, Sep Press Release, House Committee on Science, Space, & Technology, GAO Raises Questions about Adequate Supply of Lithium-7 for Nuclear Power Reactors, Oct 9,

4 EPRI Li-7 Strategy Co-funding from DOE in 2015 and 2016 Usage Reduction and Plant Impacts Summarize impacts on plant Establish methods to reduce cycle usage Document in white paper Survey of industry usage White paper Assesses the impact of Li-7 supply loss for a typical PWR plant and options for mitigating the impact Evaluation of Lithium Addition on Plant Startup EPRI report Lithium Recovery Continue to evaluate recovery/recycle options End goal full scale demonstration Develop technology for Li-7 recovery from spent resin Demonstration Li-7 recovery for industry Alternative for ph Control KOH Feasibility Gap Assessment KOH Materials/Fuels Evaluations Incorporation of KOH into MULTEQ Feasibility analysis of KOH use ( ) and gaps identified High temperature KOH Chemistry for MULTEQ and ph calculation Literature review and experimental scope development for impact on Zircaloy cladding 4

5 Feasibility of KOH vs LiOH for PWR Primary ph Control Published October 2015 ( ) Important differences between VVER and Western-PWR experiences Materials: Titanium-stabilized SS (VVER) vs nickel-based alloys (PWR) Fuel cladding: Both zirconium alloy (KOH less corrosive), but low crud and lower boiling (VVER) Chemistry: Ammonia for hydrogen (VVER) vs dissolved hydrogen gas (PWR), Li/K new to PWRs Worker dose & Radwaste: Potassium activation products (VVER) Key Gaps Materials Fuels Chemistry Radiation Safety & Radwaste SCC of austenitic SS reactor internals & pressure boundary (including IASCC) SCC initiation and CGR of nickel based alloys Corrosion and/or hydriding of zirconium fuel cladding with crud and boiling Management of Li/K ratio (e.g., ph T control, resin management) Including Li-7 production rate B-10 Li-7 Activation of Potassium and impurities (i.e., sodium) 42 K external dose 40 K internal dose Waste classification Appears feasible. Initiated next steps. More detailed multi-year plan developed. 5

6 Historical Use of KOH KOH used for ph control at Trino Vercellese (Northern Italy) Westinghouse 270 MWe PWR Operated from 1964 to 1988 Fuel clad and SG tubing was stainless steel However, very little to no data is now available KOH used in Russian-designed VVERs* * VVER = WWER = Water-Water Energetic Reactor, i.e., a PWR 6

7 Observations from VVER Operation General Observations Successful use of KOH for over 40 years Generally low corrosion No observed Crud Induced Power Shift (CIPS) Very low radiation fields No unique waste or radiation field issues Challenges Alloys are somewhat different Higher fuel boiling duty in PWRs Management of Li-7 production on ph T is a known challenge VVER experience indicates it can be managed Activation pathways of potassium Other chemistry differences Ammonia added for ECP control instead of hydrogen VVER 1000 Primary Circuit 7

8 VVER Overview ph Parameter Operated in former USSR and Eastern Europe (plus China and India) 58 in operation, 25 under construction Several types built Mainly VVER-440 & VVER-1000 Values During Full Power Operation VVER ph at 300 C: VVER-440: 7.1 to 7.3 VVER-1000: 7.0 to 7.2 EPRI Guidelines 7.0 at Operating Temperature Li (ppm) --- Typically 3.5 (1) K (equivalent) (ppm) 0.8 to (2) NH 3 (ppm) 5 (normally 10) diagnostic H 2 (cc/kg) 30 to to 50 1 Higher concentrations may be used for limited duration (with fuel vendor concurrence) 2 No EPRI limit. This is the molar equivalent based on 3.5 ppm Li Primary System Materials Ti-stabilized Stainless Steel Main coolant pumps, SG tubes Low Alloy Steel Main loop piping (clad with SS), RPV (usually clad with SS) Carbon Steel RPV Head and Pressurizer (clad with SS), Nozzles Fuel Cladding Zr-1%Nb Some designs also have Zr- 2.5%Nb sheath surrounding the assemblies 8

9 Materials Compatibility General Corrosion Stainless Steel From Primary Circuit of Temelin (VVER) Fe-rich Cr-rich Outer Oxide Inner Oxide Chromium Depleted Metal Base Metal Typical Oxide Structure Baffle-former bolt at Tihange (PWR) (10x Cr as outer oxide) T. Grygar and M. Zmitko, Corrosion Products Behavior Under VVER Primary Coolant Conditions, Chemistry 2002: International Conference on Water Chemistry in Nuclear Reactors Systems - Operation Optimization and New Developments, Avignon, (NPC 2002) Comparison of oxide films from VVERs and PWRs show similar structures and thicknesses Analytical Transmission Electron Microscopy (ATEM) Characterization of Stress-Corrosion Cracks in LWR- Irradiated Austenitic Stainless Steel Core Components Revision 2. EPRI, Palo Alto, CA:

10 Materials Compatibility General Corrosion Nickel Based Alloys Oxide layer is similar Corrosion product release kinetics appear similar BOREAL test loop of Alloy 690 tubing But Short term test Does not consider possible restructuring, i.e., from Li to K Effect of Boron Concentration on Alloy-690 Corrosion Product Release Rates - Results at 325 C and 285 C. EPRI, Palo Alto, CA : Typical Oxide Layer on Nickel Alloy in a PWR D. Morton, N. Lewis, M. Hanson, S. Rice, and P. Sander, Nickel Alloy Primary Water Bulk Surface and SCC Corrosion Film Analytical Characterization and SCC Mechanistic Implications, Lockheed Martin Corporation, Schenectady, NY: LM-07K022. ph T is likely controlling factor rather than ion-specific effect 10

11 Materials Compatibility SCC Stainless Steel PWSCC Chemistry has only a small effect within normal ranges Good performance of similar materials in VVERs Austenitic stainless steels (no nickel alloys) Irradiation Assisted Stress Corrosion Cracking (IASCC) Time to failure decreases when exposed to higher Li concentrations EDF work (limited to ten specimens, and 2.2 vs 3.5 ppm Li) EPRI MRP preliminary work Evaluate effect of potassium on SCC of stainless steel (including IASCC) O-ring Uniaxial Constant Load (UCL) EPRI MRP Preliminary Testing (at 340 C) Li = 2.0 or 8.0 ppm ph 300 C = 7.2 (Boron) Irradiated specimens at ~ 60 or 100 dpa 21 specimens tested 11

12 Materials Compatibility PWSCC Nickel Based Alloys Li has little to no effect on initiation (Metastudy) However, two individual studies (considering smaller concentration ranges) have reached a different conclusion About 40% reduction in time to initiation going from about 2.2 to 3.5 ppm Li Initiation Testing CGR Testing Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7. EPRI, Palo Alto, CA: Li has no measurable effect on crack growth rate Constant ph T 2 ppm Li 7 ppm Li 2 ppm Li 0.3 ppm Li Materials Reliability Program: Effects of B/Li/pH on PWSCC Growth Rates in Ni-Base Alloys (MRP-217). EPRI, Palo Alto, CA: Evaluate effect of potassium on PWSCC Initiation of nickel based alloys 12

13 Summary of Needed Materials Testing for Qualification of KOH Trial Application* Crack Initiation Testing Non-irradiated testing Alloy 600 (or weld metal) Cold work (at or near yield stress) 1 material, 3 chemistries (including crevice chemistry) Stainless steel Sensitized, cold work 1 material, 1 chemistry ( crevice chemistry) Irradiated testing Stainless steel (similar to current MRP testing with Li) Crack Growth Rate Testing using On the fly DCPD technique Reference LiOH #1 KOH-chemistry #1 Reference LiOH #2 KOH-chemistry #2 Reference LiOH #3 KOH-chemistry #3 Reference LiOH #4 KOH-chemistry #4 Non-irradiated testing Alloy 600 MA or SA and 182 Stainless steel (CW and sensitized) LAS Irradiated testing Stainless steel *Thank you to Peter Chou (EPRI) for developing the detailed materials testing plan 13

14 Fuel Cladding Overview Long history of good performance of Russian alloys in VVERs Recent history of good performance of Western alloys in VVERs Westinghouse supplied fuel in Ukraine and Czech Republic However, for VVERs: Boiling duties not particularly high Deposit loading is typically much lower No zinc Western Russian Fuel Type Zircaloy-2 Zircaloy-4 Zirlo M5 Zr-1Nb Zr-2.5Nb Element Wt% Wt% Wt% Wt% Wt% Wt% Nb Sn <0.01 Fe <0.05 Cr <0.015 Ammonia is another chemistry difference Ni O (Table adapted from and ) 14

15 Fuel Cladding Solubility impact on CIPS Precipitation of alkali-borate and alkali-borate-nickel compounds are a concern for inducing Crud Induced Power Shift (CIPS) Not known whether this will be an issue for KOH Very likely that potassium compounds are much more soluble Developing MULTEQ entries for important species Currently under development MULTEQ entries for KOH and K-borate system (soluble and solid species) to be incorporated in 2016 K+ Na+ Li+ mol/kg ppm mol/kg ppm mol/kg ppm (BO 2 ) , , ,325 (B 4 O 7 ) -2 (1) , ,000 (B 5 O 8 ) , ,000 Notes: (1) - Solubility is for K 2 B 4 O 7 and Na 2 B 4 O 7 15

16 Corrosion Rate (mg/dm 2 ) Corrosion Rate (mg/dm 2 ) Fuel Cladding Corrosion and Hydriding Fuel cladding corrosion expected to be much smaller with KOH chemistry Although increased solubility could lead to much higher cation concentrations at the fuel cladding surface when significant crud is present Dominant source of hydrides in zirconium based alloys are from hydrogen released as a result of the cladding corrosion process Hydrogen pick up fraction (HPUF) Fraction of corrosion generated hydrogen that is absorbed into the cladding NaOH LiOH KOH NaOH LiOH KOH Zircaloy 2 H. Coriou, L. Grall, J. Neunier, M. Pelras, and H. Willermoz, The Corrosion of Zircaloy in Various Alkaline Media at High Temperature, Corrosion of Reactor Materials, Vol. II, 193, IAEA, Vienna (1962). 10 Corrosion Rate of Zircaloy 2 at 360 C Concentration of Cations (ppm) Concentration of Cation (ppm) Y.H. Jeong, J.H. Baek, S.J. Kim, H.G. Kim, and H. Ruhmann, Corrosion Characteristics and Oxide Microstructures of Zircaloy-4 in Aqueous Alkali Hydroxide Solutions, Journal of Nuclear Materials 270:3, Corrosion and HPUF is expected to be lower with KOH than with LiOH 16

17 Potassium concentration, ppm Lithium concentration, ppm Chemistry Challenge ph T Control Li-7 generation expected to be comparable to VVERs B( n, ) 10 7 Li Needs to be calculated for specific cycle/core designs Boric acid concentration, g/l 4 Potassium Lithium 5 6 Potassium Hydroxide: A Potential Mitigation for AOA. EPRI, Palo Alto, CA: TE Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7. EPRI, Palo Alto, CA: Detailed analysis needed Actual PWR generation rates of 7 Li Potassium addition rates (ph T program) Selectivity of resins Effect of ammonia Demineralizer capacity Simultaneous control of Li and K required. This work is underway. 17

18 Radiation Safety & Radwaste Challenges For reactions reviewed, 42 K is the candidate most likely to be problematic High energy gamma (1.5 MeV), half-life of 12.3 hours VVER operating experience indicates 42 K is a significant contributor to radiation fields during operation Could be significant benefits to tightening specifications on impurities in bulk KOH (e.g., Na) Worker dose and waste issues must be addressed Internal dose from 40 K (now that it s produced by reactor operation) Impact on waste classification and disposal Activation pathways and production of important isotopes needs to be addressed Isotope Absorption Cross Section (barn) Natural Abundance 6 Li Li Li (1) K K K K (1) Na (2) Ca 0.68 (3) Notes (1) Average cross section based on natural abundance (2) 23 Na is the only stable isotope of sodium (3) 42 Ca is produced through the activation of 41 K Some Nuclear Reactions to Consider: K( n, ) K K( n, p) Ar K( n, ) K K( n, p) Ar K( n, ) K K Ca Na( n, ) Na Na( n, p) Ne 18

19 Qualification Plan Shortest Timeline 19

20 Together Shaping the Future of Electricity 20