Thermal Annealing of Reactor Pressure Vessels Is a Needed ... · PDF filepipe supports,...

November 8, 2013

IAEA, Vienna, Austria

William Server & Mikhail Sokolov ATI Consulting ORNL

Degradation of Primary Components of Pressurized Water Cooled Nuclear Reactors: Current Issues and Future Challenges

Thermal Annealing of Reactor Pressure Vessels Is a Needed Mitigation Option

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Thermal Annealing Generally is Considered a Last Resort Embrittlement Mitigation Method

• Other mitigation methods are preferred?

• Annealing methodology is unproven?

• Cost?

• May damage other parts of plant?

• Violates original vessel design conditions/properties?

• Requires new surveillance program?

• Relicensing and license extension issues?

• Public intervention/non-trust?

• Political issues?

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BACKGROUND AND HISTORY

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What is or is not RPV Thermal Annealing?

• It is not a traditional metallurgical anneal at temperatures up to about 1000oC

• It is a localized area heat treatment of the RPV (at lower temperatures between 340oC and 500oC) and for a long holding time (up to 168 h)

• Material mechanical properties are partially or fully recovered, but resulting fine scale microstructure is different

• Re-irradiation response can be different due to the different starting microstructure

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Vessel Annealing

• Thermal annealing is the only option that can, to some degree, recover irradiated beltline region transition temperature shift and recover upper shelf energy properties lost during radiation exposure and extend RPV service life

• Beltline region is heated to 340 to 500°C

– Amount of mechanical property recovery is function of:

• Difference between the irradiation and thermal anneal temperatures

• Time of annealing

• Material chemistry

• Degree of pre-existing irradiation damage

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Irradiation Damage / Annealing Recovery

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Vessel Annealing Process

• Selection of anneal temperature requires some trade-offs

– Higher temps (and longer annealing times) produce greater recovery of fracture toughness and tensile strength properties, but may stimulate thermal aging

– Higher temps also pose greater engineering challenges to assure that the annealing operation does not distort or damage the vessel, supports, primary coolant piping, pipe supports, adjacent concrete, insulation, etc.

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Vessel Annealing – Wet Anneal

• Two basic types of anneals

– Wet anneal

– Dry anneal

• Wet anneal is performed at temperatures < 650°F (343°C)

– Reactor coolant water is generally heated by the RCPs

– Wet annealing is not as complicated from an engineering viewpoint because primary water temperature is controlled by pump heat up to the vessel design temperature of 343°C

• Wet anneals have been successful on two test reactors, SM-1A (US Army, Alaska, 1967) and BR3 (Belgium, 1984); they operated near 260°C for a short time after

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Wet Annealing

• Because the normal operating temperature of many commercial US PWRs is approximately 288°C, wet annealing is only partially effective

– Temperature differential between normal operating temperature and 343°C annealing temperature is not high enough to obtain substantial mechanical property recovery

• Wet annealing is not a practical solution for current operating PWRs

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Dry Annealing

• Dry anneals are performed at higher temperatures than wet anneals

– Use air as the heating medium inside of radiant can

– Electric-resistance heating source

• Dry annealing requires removal of core internal structures and primary water so that a radiant heating source can be inserted near the vessel wall to locally heat the embrittled beltline region

– Engineering difficulties of dry anneal process are quite complex and may need plant-specific evaluations to assure that other portions of the plant (eg., concrete) are not harmed by the high annealing temperatures

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DRY ANNEALING EXPERIENCE

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VVER-440 RPV Thermal Annealing

• The technology of thermal annealing of the VVER-440 reactors had been developed in Soviet Union in the first half of 1980s and first implemented in 1987

• Thermal annealing was viewed as the last resort in mitigating embrittlement of RPVs

• More preferable options, at that time, were:

– Low-leak core design

– Pre-heating of emergency coolant to 55ºC

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Summary of VVER-440 RPV Thermal Annealing

• The technology for thermal annealing was developed for Soviet VVER-440 reactors in first half of 1980s

• The “dry” annealing method was selected

• Fifteen VVER-440 reactors have been annealed in Russia and Eastern Europe from 1987 to 2010 and one reactor has been annealed twice

• Four reactors in Russia, one in Finland, and one in Ukraine are still in operation and all will exceed 40-year term

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Vessel Annealing Recovery Results in US

• Annealing recovery test results on US RPV steels found:

– Annealing at 850°F (454°C) resulted in complete recovery of USE, and 75% or more recovery of the Charpy 41 J transition temperature shift

– Annealing at 343°C provided significantly less benefit

– EPRI report TR-106001, Dec. 1995 reported results of some irradiation embrittlement and re-annealing studies for Yankee Rowe related materials

• Annealing at 454°C resulted in recovery of 80-100% of the transition temperature and 100% recovery of the USE

• Annealing at 343°C resulted in about a 40% recovery in transition temperature

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Dry Annealing Evaluations

• 454°C is regarded as an optimum “dry” annealing temperature

• In 1980s a study at Idaho National Laboratory assessed annealing feasibility for US RPVs including alternative heating methods

• In 1995 a study was conducted to determine if thermal annealing of the reactor vessel in Westinghouse 3 and 4 loop plants is feasible

– Thermal and stress analyses determined that stress, temperature and dimensions of the vessel and its associated components remain within acceptable limits

– Conclusion: there are no major technical impediments to thermal annealing the vessels studied

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Marble Hill Demonstration Project

• In 1990’s a joint DOE/industry-sponsored Annealing Demonstration Project (ADP) was conducted at the Marble Hill facility (a partially completed Westinghouse plant) to demonstrate feasibility

– Nozzle-supported four loop Westinghouse design vessel -- canceled plant (unirradiated vessel)

– Indirect gas-fired heating method was chosen

– DOE funding lost after demonstration was completed, but EPRI funded writing the Marble Hill report (EPRI TR-104934) and a final NRC report (NUREG/CR-6552) was also later published

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Marble Hill Demonstration Project

• Objectives

– Demonstrate engineering feasibility of annealing system

– Determine component thermal/stress response

• Determine RPV dimensional stability following anneal

• Biological shield wall temperature

• Magnitude of thermally induced stresses in nozzle region (satisfy ASME Code Case N-557)

– Benchmark 3-D stress analysis models

– Provide information to resolve regulatory concerns

– Provide realistic cost data regarding equipment costs, modeling, etc.

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Midland Demonstration Project

• Skirt-supported Babcock & Wilcox-design vessel

• Electric resistance heating method (Russian technology and experience)

• Project approximately 50 percent complete when DOE funding eliminated

• Electric resistance heater fabricated and tested, but never shipped to the US from Russia

• Demonstration never completed

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Why the Demonstration? – Palisades NPP

• Palisades was limited to operate until 1999 based upon PTS concerns for the most-limiting weld metal heat (W5214 axial welds); other welds: 27204 and 34B009

• Material chemistry variability issue as a result of sampling of retired steam generators containing two of the same key welds (W5214 and 34B009)

• Planned to anneal in 1998 to recover properties and continue operation to at least 2011 and hopefully beyond

• Developed a supplemental surveillance program to assess material recovery and re-embrittlement trends for all beltline welds and the surveillance plate material

• Annealing canceled due to revised fluence estimates; also concern about public hearings when authorized to anneal

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US RULES, REGULATIONS, AND GUIDANCE

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NRC Regulations/Guidance

• Annealing Rule in 10 CFR Part 50.66

• Regulatory Guide 1.162 on Annealing Program requirements and reporting

– Describes the format and content of an acceptable Thermal Annealing Report (TAR)

• Annealing Rule and Regulatory Guide 1.162 contain reference to NUREG/CR-6327

– Predictive model for annealing recovery utilizing microhardness and CVN data to cover a broad range of annealing conditions

– Model incorporates annealing time and temperature and neutron irradiation dose rate (flux)

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Effect of Dose Rate on Annealing Recovery

• NUREG/CR-6327 model illustrates a strong annealing recovery effect due to the initial irradiation dose rate for annealing temperatures less than 427oC

• For most “dry annealing” considerations the effect of dose rate on annealing recovery is irrelevant since the temperature is well above 427oC

• Other recent data from an IAEA project on VVER-440 steels support no dose rate effects (MTR vs. actual surveillance) for annealing temperatures near 460oC

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NUREG/CR-6327 Predicted Recovery

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Guidance for Thermal Annealing

• 10 CFR 50.66, “Requirements for thermal annealing of the reactor pressure vessel”

– Permits thermal annealing of LWRs

– Requires a plan for conducting the thermal annealing be submitted at least three years before fracture toughness criteria are exceeded

– Reg Guide 1.162 describes the format and content of an acceptable Thermal Annealing Report (TAR) and addresses the metallurgical and engineering issues that need to be addressed in an application to perform a thermal annealing

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ASTM Activities

• ASTM E 509 revised in 1997 to provide expanded guidance on thermal annealing and the necessary supplemental material surveillance programs; ASTM E 509 was further updated in 2003

• Splitting of old ASTM E 185 into two new Practices (ASTM E 185-02 on Surveillance Program Design and ASTM E 2215-02 on Testing of Surveillance Capsules); both emphasize and encourage fracture toughness Master Curve testing in addition to CVN testing

• Small specimen test techniques may be applied to annealing applications using other ASTM guides

– ASTM E 636 on Supplemental Test Techniques

– ASME E 1253 on Charpy Specimen Reconstitution

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Potential for Intergranular Fracture – Caution Stated in ASTM E 509

• Phosphorus segregation to grain boundaries in ferritic steels occurs during irradiation and thermal annealing sequences, which can lead to non-hardening intergranular fracture; key experimental examples include: coarse-grained HAZ, forging “ghost lines” areas, and non-standard heat treatments (akin to temper embrittlement)

• For U.S. materials (with low P content), the degree of phosphorus segregation does not appear to create any significant intergranular fracture potential, but some intergranular fracture areas are evident on some fracture surfaces (even in non-irradiated RPV steels)

• For higher P steels, this issue related to non-hardening embrittlement needs to be adequately assessed for long-term operation after annealing

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ASME Code Case N-557

• “In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1)”

• Provides Code guidance for assuring design conformance after performing a thermal anneal heat treatment

– Limits magnitude of thermally induced stresses in nozzle region

– Effectively limits the maximum temperature of annealing to 505oC

– Passed in 1995 in anticipation of Palisades NPP thermal anneal

• Technical basis published by EPRI in TR-106967

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Re-Embrittlement is Key Issue in ASTM E 509 and NRC Guidelines

• ASTM E 509 provides detailed guidance for a supplemental surveillance program to monitor reirradiation embrittlement

• Annealing database was compiled in the mid-late 1990s:

– Annealing recovery generally follows NUREG/CR-6327 model

– Re-embrittlement falls between the lateral shift and vertical shift models (mechanistically related to tie-up of copper in clusters from original irradiation which then over-ripen)

– EPRI computer code allows prediction based on a mix of lateral and vertical shift – initial attempts at supporting percentage based on mechanistic understanding were made

– Dose rate effects must be considered for reirradiation, similar to current surveillance concerns for initial irradiation

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Methods Used To Predict Re-embrittlement

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Key Annealing Issues as Related to Long Term Operation

• Based on NRC Regulations and guidance – evidence of dose rate on annealing recovery at low annealing temperatures (less than 427oC)

• Based on guidance in ASTM E 509

– Re-embrittlement rate and surveillance during extended life (including any effect of dose rate)

– Potential enhancement of P segregation and intergranular fracture

• Based on ASME Code Case – must minimize thermally induced stresses in nozzle region, which effectively limits maximum temperature of annealing to 505oC

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MECHANISTIC UNDERSTANDING

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Thermal Annealing Knowledge Related to Mechanisms

• As with irradiation damage mechanisms that cause RPV material embrittlement, the mechanisms of thermal annealing that result in recovery of the material toughness are quite complicated

• For a copper-bearing material such as Linde 80 welds, irradiation of the material creates copper-rich precipitates (CRPs), which are about 1-2 nm in diameter, and various forms of matrix damage (e.g., dislocation loops)

• The thermal annealing treatment dissolves most of the small irradiation-induced defect clusters and complexes, but the dissolution of the CRPs actually results in an increase in size to 3- 5nm, slightly reduces the volume fraction, and significantly reduces the number density

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Multiple Anneals

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SANS and APT

• Available Small Angle Neutron Scattering (SANS) and Atom Probe Tomography (APT) data on irradiation/annealing/reirradiation of copper bearing materials can be summarized as follows:

– In the irradiated condition, a high number density of ultrafine Cu-Mn-, Ni- and Si-enriched precipitates were observed

– In the irradiated/annealed condition, the number density of Cu-enriched precipitates had decreased more than an order of magnitude, and their size (radius) had significantly increased.

– In the irradiated/annealed/reirradiated condition, a high number density of ultrafine Cu, Mn, Ni and Si-enriched precipitates were observed

• Very similar to the irradiated condition

• Additionally, solute segregation of Cu, Ni, Mn, Si, and P to dislocations was observed

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APT Results as Irradiated

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APT Results after Annealing

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Concluding Remarks

• Technology exists and has been proven for performing thermal annealing on VVER-440 and PWR RPVs in the beltline region

• Decision to anneal may involve more than technical or cost issues

– When annealing is technically warranted, it should be seriously considered and not discarded as a “last resort”

– Even if thermal annealing is not technically needed, a decision to anneal could benefit extended long term operation

• Guarantee that the fracture toughness properties are improved

• Re-embrittlement rate should be reduced

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