Fuel Cladding
Potassium Hydroxide for PWR Primary Coolant pHT 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 © 2016 Electric Power Research Institute, Inc. All rights reserved.
PWRs Currently Need Highly Enriched Li-7 pHT 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
63𝐿𝑖 + 𝑛 → 42𝐻𝑒 + 31𝐻 – Need enriched Li-7 (99.99%) Li-7 is generated in coolant via neutron reaction with B-10 –
10 5𝐵
+ 𝑛 → 73𝐿𝑖 + 𝛼
2 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Natural Lithium
Isotope
Abundance (at%)
Li-6
7.59
Li-7
92.41
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 supply1 – “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-13-716, “Managing Critical Isotopes: Stewardship of Lithium-7 Is Needed to Ensure a Stable Supply”, Sep. 2013.
2
Press Release, House Committee on Science, Space, & Technology, “GAO Raises Questions about Adequate Supply of Lithium-7 for Nuclear Power Reactors”, Oct 9, 2013.
3 © 2016 Electric Power Research Institute, Inc. All rights reserved.
EPRI Li-7 Strategy Co-funding from DOE in 2015 and 2016 Survey of industry usage White paper
Usage Reduction and Plant Impacts • Summarize impacts on plant • Establish methods to reduce cycle usage • Document in white paper
Lithium Recovery • Continue to evaluate recovery/recycle options • End goal – full scale demonstration
Alternative for pH Control • KOH Feasibility Gap Assessment • KOH Materials/Fuels Evaluations • Incorporation of KOH into MULTEQ
–
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 3002008184
Develop technology for Li-7 recovery from spent resin Demonstration Li-7 recovery for industry
Feasibility analysis of KOH use (3002005408) and gaps identified High temperature KOH Chemistry for MULTEQ and pH calculation Literature review and experimental scope development for impact on Zircaloy cladding
4 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Feasibility of KOH vs LiOH for PWR Primary pH Control Published October 2015 (3002005408) 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 • SCC of austenitic SS reactor internals & pressure boundary (including IASCC) • SCC initiation and CGR of nickel based alloys
Fuels
Chemistry
Radiation Safety & Radwaste
• Corrosion and/or hydriding of zirconium fuel cladding – with crud and boiling
• Management of Li/K ratio (e.g., pHT control, resin management) • Including Li-7 production rate B-10 Li-7
• Activation of Potassium and impurities (i.e., sodium) • 42K – external dose • 40K – internal dose • Waste classification
Appears feasible. Initiated next steps. More detailed multi-year plan developed. 5 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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
VVER 1000 – Primary Circuit
– Alloys are somewhat different – Higher fuel boiling duty in PWRs – Management of Li-7 production on pHT 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 7 © 2016 Electric Power Research Institute, Inc. All rights reserved.
VVER Overview Parameter
pH Li (ppm) K (equivalent) (ppm)
Values During Full Power Operation VVER
EPRI Guidelines
pH at 300°C: VVER-440: 7.1 to 7.3 VVER-1000: 7.0 to 7.2
≥ 7.0 at Operating Temperature
---
Typically ≤ 3.5 (1)
0.8 to 20
Primary System Materials – Ti-stabilized Stainless Steel Main coolant pumps, SG tubes
– Low Alloy Steel ≤ 19.55
(2)
NH3 (ppm)
≥ 5 (normally 10)
diagnostic
H2 (cc/kg)
30 to 60
25 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
Operated in former USSR and Eastern Europe (plus China and India) – 58 in operation, 25 under construction
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 Zr2.5%Nb sheath surrounding the assemblies
Several types built – Mainly VVER-440 & VVER-1000 8 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Materials Compatibility General Corrosion – Stainless Steel
Fe-rich
Outer Oxide
Cr-rich
Inner Oxide
Chromium Depleted Metal
From Primary Circuit of Temelin (VVER)
Typical Oxide Structure Base Metal
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, 2002. (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 LWRIrradiated Austenitic Stainless Steel Core Components—Revision 2. EPRI, Palo Alto, CA: 2006. 1014511.
9 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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…
Typical Oxide Layer on Nickel Alloy in a PWR
•
Short term test
•
Does not consider possible restructuring, i.e., from Li to K
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: 2007. LM-07K022.
pHT is likely controlling factor rather than ion-specific effect
Effect of Boron Concentration on Alloy-690 Corrosion Product Release Rates - Results at 325°C and 285°C. EPRI, Palo Alto, CA : 2001. 1011744. 10 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Uniaxial Constant Load (UCL)
Materials Compatibility SCC – Stainless Steel
O-ring
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
EPRI MRP Preliminary Testing (at 340°C)
• Li = 2.0 or 8.0 ppm • pH300°C = 7.2 (Boron)
Evaluate effect of potassium on SCC of stainless steel (including IASCC) 11 © 2016 Electric Power Research Institute, Inc. All rights reserved.
• Irradiated specimens at ~ 60 or 100 dpa • 21 specimens tested
Materials Compatibility
Initiation Testing
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
CGR Testing Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7. EPRI, Palo Alto, CA: 2013. 3002000505.
Li has no measurable effect on crack growth rate –
Constant pHT 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: 2007. 1015008.
Evaluate effect of potassium on PWSCC Initiation of nickel based alloys
12 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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 Reference LiOH #2 Reference LiOH #3 Reference LiOH #4
KOH-chemistry #1 KOH-chemistry #2 KOH-chemistry #3 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 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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 “Western”
However, for VVERs: – Boiling duties not particularly high
Fuel Type
Zircaloy-2
Zircaloy-4
Zirlo®
M5®
Zr-1Nb
Zr-2.5Nb
Element
Wt%
Wt%
Wt%
Wt%
Wt%
Wt%
0.94-0.98
1.0
1.0
2.4-2.8
0.06
0.09-0.13
Nb
– Deposit loading is typically much lower – No zinc – Ammonia is another chemistry difference
Russian
Sn
1.2-1.7
1.2-1.7
0.97
Co-authors: Chuck Marks and Jeff Reinders Dominion Engineering, Inc.
Joel McElrath, Daniel M. Wells, Paul Frattini, Al Ahluwalia, Ryan Wolfe EPRI © 2016 Electric Power Research Institute, Inc. All rights reserved.
PWRs Currently Need Highly Enriched Li-7 pHT 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
63𝐿𝑖 + 𝑛 → 42𝐻𝑒 + 31𝐻 – Need enriched Li-7 (99.99%) Li-7 is generated in coolant via neutron reaction with B-10 –
10 5𝐵
+ 𝑛 → 73𝐿𝑖 + 𝛼
2 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Natural Lithium
Isotope
Abundance (at%)
Li-6
7.59
Li-7
92.41
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 supply1 – “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-13-716, “Managing Critical Isotopes: Stewardship of Lithium-7 Is Needed to Ensure a Stable Supply”, Sep. 2013.
2
Press Release, House Committee on Science, Space, & Technology, “GAO Raises Questions about Adequate Supply of Lithium-7 for Nuclear Power Reactors”, Oct 9, 2013.
3 © 2016 Electric Power Research Institute, Inc. All rights reserved.
EPRI Li-7 Strategy Co-funding from DOE in 2015 and 2016 Survey of industry usage White paper
Usage Reduction and Plant Impacts • Summarize impacts on plant • Establish methods to reduce cycle usage • Document in white paper
Lithium Recovery • Continue to evaluate recovery/recycle options • End goal – full scale demonstration
Alternative for pH Control • KOH Feasibility Gap Assessment • KOH Materials/Fuels Evaluations • Incorporation of KOH into MULTEQ
–
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 3002008184
Develop technology for Li-7 recovery from spent resin Demonstration Li-7 recovery for industry
Feasibility analysis of KOH use (3002005408) and gaps identified High temperature KOH Chemistry for MULTEQ and pH calculation Literature review and experimental scope development for impact on Zircaloy cladding
4 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Feasibility of KOH vs LiOH for PWR Primary pH Control Published October 2015 (3002005408) 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 • SCC of austenitic SS reactor internals & pressure boundary (including IASCC) • SCC initiation and CGR of nickel based alloys
Fuels
Chemistry
Radiation Safety & Radwaste
• Corrosion and/or hydriding of zirconium fuel cladding – with crud and boiling
• Management of Li/K ratio (e.g., pHT control, resin management) • Including Li-7 production rate B-10 Li-7
• Activation of Potassium and impurities (i.e., sodium) • 42K – external dose • 40K – internal dose • Waste classification
Appears feasible. Initiated next steps. More detailed multi-year plan developed. 5 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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
VVER 1000 – Primary Circuit
– Alloys are somewhat different – Higher fuel boiling duty in PWRs – Management of Li-7 production on pHT 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 7 © 2016 Electric Power Research Institute, Inc. All rights reserved.
VVER Overview Parameter
pH Li (ppm) K (equivalent) (ppm)
Values During Full Power Operation VVER
EPRI Guidelines
pH at 300°C: VVER-440: 7.1 to 7.3 VVER-1000: 7.0 to 7.2
≥ 7.0 at Operating Temperature
---
Typically ≤ 3.5 (1)
0.8 to 20
Primary System Materials – Ti-stabilized Stainless Steel Main coolant pumps, SG tubes
– Low Alloy Steel ≤ 19.55
(2)
NH3 (ppm)
≥ 5 (normally 10)
diagnostic
H2 (cc/kg)
30 to 60
25 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
Operated in former USSR and Eastern Europe (plus China and India) – 58 in operation, 25 under construction
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 Zr2.5%Nb sheath surrounding the assemblies
Several types built – Mainly VVER-440 & VVER-1000 8 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Materials Compatibility General Corrosion – Stainless Steel
Fe-rich
Outer Oxide
Cr-rich
Inner Oxide
Chromium Depleted Metal
From Primary Circuit of Temelin (VVER)
Typical Oxide Structure Base Metal
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, 2002. (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 LWRIrradiated Austenitic Stainless Steel Core Components—Revision 2. EPRI, Palo Alto, CA: 2006. 1014511.
9 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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…
Typical Oxide Layer on Nickel Alloy in a PWR
•
Short term test
•
Does not consider possible restructuring, i.e., from Li to K
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: 2007. LM-07K022.
pHT is likely controlling factor rather than ion-specific effect
Effect of Boron Concentration on Alloy-690 Corrosion Product Release Rates - Results at 325°C and 285°C. EPRI, Palo Alto, CA : 2001. 1011744. 10 © 2016 Electric Power Research Institute, Inc. All rights reserved.
Uniaxial Constant Load (UCL)
Materials Compatibility SCC – Stainless Steel
O-ring
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
EPRI MRP Preliminary Testing (at 340°C)
• Li = 2.0 or 8.0 ppm • pH300°C = 7.2 (Boron)
Evaluate effect of potassium on SCC of stainless steel (including IASCC) 11 © 2016 Electric Power Research Institute, Inc. All rights reserved.
• Irradiated specimens at ~ 60 or 100 dpa • 21 specimens tested
Materials Compatibility
Initiation Testing
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
CGR Testing Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7. EPRI, Palo Alto, CA: 2013. 3002000505.
Li has no measurable effect on crack growth rate –
Constant pHT 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: 2007. 1015008.
Evaluate effect of potassium on PWSCC Initiation of nickel based alloys
12 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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 Reference LiOH #2 Reference LiOH #3 Reference LiOH #4
KOH-chemistry #1 KOH-chemistry #2 KOH-chemistry #3 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 © 2016 Electric Power Research Institute, Inc. All rights reserved.
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 “Western”
However, for VVERs: – Boiling duties not particularly high
Fuel Type
Zircaloy-2
Zircaloy-4
Zirlo®
M5®
Zr-1Nb
Zr-2.5Nb
Element
Wt%
Wt%
Wt%
Wt%
Wt%
Wt%
0.94-0.98
1.0
1.0
2.4-2.8
0.06
0.09-0.13
Nb
– Deposit loading is typically much lower – No zinc – Ammonia is another chemistry difference
Russian
Sn
1.2-1.7
1.2-1.7
0.97
