Fusion and Spallation Irradiation Conditions Steven J

Fusion and Spallation Irradiation Conditions Steven J. Zinkle

Metals and Ceramics Division

Oak Ridge National Laboratory, Oak Ridge, TN

International Workshop on Advanced Computational Science: Application to Fusion and Generation-IV Fission Reactors Washington DC, March 31-April 2, 2004 Comparison of fission and fusion (ITER) neutron spectra Stoller & Greenwood, JNM 271-272 (1999) 57

• Main difference between fission and DT fusion neutron spectra is the presence of significant flux above ~4 MeV for fusion –High energy neutrons typically cause enhanced production of numerous transmutation products including H and He –The Primary Knock-on Atom (PKA) spectra are similar for fission and fusion at low energies; fusion contains significant high-energy PKAs (>100 keV) Displacement Damage Mechanisms are being investigated with Molecular Dynamics Simulations Damage efficiency saturates when subcascade formation occurs

Avg. Avg. fission fusion

Molecular dynamics modeling of displacement cascades up to 50 keV PKA 200 keV and low-dose experimental tests (microstructure, tensile (ave. fusion) properties, etc.) indicates that defect production from fusion and fission neutron collisions are similar 10 keV PKA => Defect source term is similar for fission and fusion conditions (ave. fission) Peak damage state in iron cascades at 100K 5 nm A critical unanswered question is the effect of higher transmutant H and He production in the fusion spectrum Radiation Damage can Produce Large Changes in Structural Materials

• Radiation hardening and embrittlement (<0.4 TM)

• Irradiation creep (<0.45 TM)

• Volumetric swelling from void formation (0.3-0.6 TM)

• High temperature He embrittlement (>0.5 TM)

In addition... • The irradiation environment associated with a D-T fusion reactor is more severe than in fission reactors – Higher lifetime dose requirements for structure – Higher He generation rates (promotes He embrittlement of grain boundaries, void swelling) Low tensile ductility in FCC and BCC metals after irradiation at low temperature is due to formation of nanoscale defect clusters Tensile ductility of neutron- irradiated Cu alloys 10 T =36-90˚C Unirradiated irr elongations Fabritsiev et al (1996)

(%) 8 Heinisch et al (1992) Singh et al. (1995)

Elongation 4

2 Uniform

0 0.0001 0.001 0.01 0.1 1 10 Damage Level (dpa) Outstanding questions to be resolved include: . Can the defect cluster formation be modified by appropriate use of nanoscale 2nd phase features or solute additions? . Can the poor ductility of the irradiated materials be mitigated by altering the predominant deformation mode?

Cu, Tirr=90˚C, 0.5 dpa Why is He/dpa ratio an important parameter for fusion materials R&D? • He generation can alter the microstructural evolution path of irradiated materials (pronounced effects typically occur for >100 appm He) – Cavity formation (matrix and grain boundaries) – Precipitate and dislocation loop formation He bubbles on grain boundaries can cause Swelling in stainless steel is maximized severe embrittlement at high temperatures at fusion-relevant He/dpa values

Grain boundary

Management of He transmutation products (matrix trapping at engineered 2nd phases) is a key factor for fusion materials Helium embrittlement of grain boundaries occurs at high temperatures for helium concentrations above ~100 appm

Creep Rupture Life of 20% Cold-worked Helium Embrittlement in Vanadium Alloys Type 316 Stainless Steel at 550˚C, 310 MPa T/TM 104 0.28 0.32 0.36 0.4 0.44 0.48 0.52 0.56 0.6 (2.0 dpa) dε/dt=0.25-1.1x10-3 s-1 1000 (no He) tot 1 100 (3.4 dpa) 2 appm He (He)/e tot 10 200 appm He

Creep rupture life (h) 25 appm He

0.1 (43 dpa)

0.01 4 0.1 0.1 0 1 10 100 1000 10 Tensile elongation ratio, e 600 700 800 900 1000 1100 1200 1300 Helium concentration (appm) Temperature (K) He trapping at nanoscale precipitates within grains is key for inhibiting He embrittlement

However…….. The formation and microstructural stability of these precipitates is strongly affected by irradiation parameters, in particular the He/dpa ratio Swelling Resistant Alloys For Fission Reactors Were Successfully Developed 14

Tirr =400-500˚C (%) 12

10 316 stainless steel 8 Swelling Ti-modified 6 316 stainless steel 4 2 Ferritic steel Volumetric 0 0 50 100 150 200 Damage Level (dpa)

• Lowest swelling is observed in body-centered cubic alloys (V alloys, ferritic steel) • Materials science strategy used for developing swelling-resistant stainless steel can be applied to new alloys if effects of He on void swelling are understood SwellingSwelling behaviorbehavior ofof stainlessstainless steelssteels

Austenitic SS Ferritic/martensitic SS

Y. Katoh et al., 2003

 Swelling behavior of RAFMs and austenitics near peak-swelling temperatures are similar in the presence of helium, except for incubation dose. Cavity formation in JLF-1 by dual-ion beam irradiation 3+ + － 6.4 MeV Fe ＋ Energy-Degraded 1.0 MeV He , 20 dpa at 743 K － Y. Katoh et al., 2003 0 ]

m Range of helium deposition µ 1

LB LB 2 Energy-Degraded 1.0MeV He+ 6.4MeV Fe3+

3 Void- free

Distance from Irradiated Surface [ zones

4 Fusion materials research must rely heavily on modeling due to inaccessibility of fusion-relevant operating regime • Extrapolation from currently available parameter space to fusion regime is much larger for fusion materials than for plasma physics program – Most of He effects data on irradiated materials is based on austenitic stainless steel (FCC); relevance to BCC alloy systems is uncertain – Recent fusion materials R&D has focused on low-dose deformation and fracture issues • An intense neutron source such as IFMIF is proposed to develop and qualify fusion structural materials prior to a construction decision for a fusion Demo reactor

Summary of Helium and Dose Parameter Range Investigated by the Fusion Materials Program 21 10 10000 Ignition s) RTNS-II FFTF -3 20 Breakeven DHCE-V alloy 10 HFIR Ni-doped F/M steel

(m ORR/HFIR spectral tailor 1980-1994 fusion τ 1000 HFIR isotopic tailor steels e 1990s HFIR target/RB 316 SS (316SS) reactor n 1019 1980s 100 1975-80 1018 He Quality, 1970-75

1017 1965 appm 10 1983-1992 (FFTF) 1016 1960 Magnetic Inertial 1 1980-1986 Confinement 1955 (RTNS-II) 1995-2001 1015 105 106 107 108 109 0.1 Ion Temperature (K) 0.01 0.1 1 10 100 1000 displacement damage (dpa) Spallation Sources

Transmutation of Neutrons as probes Tritium production nuclear waste for condensed matter ADS (F, B...) ISIS (GB) TRISPAL (F) ATW/AAA (US) LANSCE (US) APT (US) EA (I, E, F) SINQ (CH) SNS (US) JPARC (J) ESS (EU) Typical Parameters: Beam power several MW (~1 GeV protons) Pulse length ~ 1µs (several 1014 p/Pulse  100 kJ) Repetition frequency 50/60 Hz Thermal n-flux up to 7x1018 n/m2⋅s average up to 2x1021 n/m2⋅s in pulse Design of the SNS

Target Module

Outer Reflector Plug

Target Inflatable seal

Core Vessel water cooled shielding Proton Core Vessel Beam Multi-channel flange Overlap in Temperature for Fusion, Generation IV Fission Reactors and Spallation Facilities L.K. Mansur et al. ICFRM-11, J.Nucl. Mater. in press (2004) Operating Temperatures and Radiation Effects

Gen IV SCWR NGNP (VHTR)

ITER A-SSTR2 Fusion DEMO SS Temp. Trans. Spallation SNS Limit Temperature, C

Example: Austenitic SS 0 500 1000 1500 Swelling Irradiation Creep

High T He Embrittlement Low T Embrittlement (Self-defects, He, H) Key Operating Conditions for Structural Materials

Fusion Fission Spallation (Gen IV)

Coolant H2O, He, Li, PbLi, H2O(SC), He, Pb, Hg, PbBi, H2O FLiBe PbBi

Particle Energy < 14 MeV < 1 - 2 MeV < 1 GeV (p and n) Temperatures 300-1000˚C 300-1000˚C 100-500˚C

Max displacement ~200 30-150 ~20 damage He/dpa 10 appm/dpa ~0.1 appm/dpa 100 appm/dpa

Stresses Moderate, nearly Moderate, nearly High, pulsed constant constant

Based on L.K. Mansur et al. ICFRM-11, J.Nucl. Mater. in press (2004) Conclusions • The structural materials response to exposure to fission, fusion and spallation neutron environments… • Offers many similarities: – Similar defect production source term

– Radiation hardening and embrittlement (<0.4 TM) – Irradiation creep (<0.45 TM) – Phase stability, radiation induced segregation phenomena (0.3-0.6 TM)

• And some important differences: – Potential for He-enhanced radiation hardening and embrittlement (<0.4 TM) – He and H effects on void swelling (0.3-0.6 TM) – High temperature He embrittlement (>0.5 TM) – He-modified phase stability (0.3-0.6 TM) Backup viewgraphs Fusion Materials Irradiation Facilities Radiation stability is strongly dependent on exposure temperature, displacement damage (dpa), damage rate, solute transmutation (H, He, ...) => He/dpa ratio is a useful radiation effects metric for fusion materials

Comparison of He and Damage Production Rates for • Fission sources Fission, Fusion-relevant and Spallation Neutron Facilities fusion-relevant displacement damage 1000 low He generation (except Ni alloys, etc.) Ferritic Steel IFMIF-HF • Ion accelerators 100 PIREX IFMIF-MF DEMO generally limited to microstructural studies 2.5MW/m2 typically very high damage rates SNS beam dump (p) SNS • Spallation sources below target (n) 10 HFIR-target • D-Li stripping source FFTF RTNS-II

HFIR-RB 1 LANSCE ATR-ITV (n only)

HFR He Production Rate (appm He/year) 0.1 0.1 1 10 100 Displacement Damage Rate (DPA/year) under planning: IFMIF

Modified from H. Bolt (IPP-Garching), 2002 SNS Site Layout

Ring Injection Dump Concept 3

New Test Facility Concepts Concept 4 1, 2

SNS Experimental Facilities Oak Ridge Specific Issues of Common Interest

• No prototype facilities available for Fusion, Gen IV Reactors, or Liquid Metal Pulsed Targets • All irradiations conducted in few facilities, e.g HFIR and ATR JMTR and JOYO HFR BOR 60 LANSCE SINQ • Common alloys--austenitics, ferritic-martensitics, high nickel alloys, mechanically alloyed steels • Theory and modeling of radiation response • Can one technology provide radiation source for other technologies-- e.g., spallation for fusion, IFMIF for Gen IV, Gen IV for Fusion • High transmutation gas effects on swelling and embrittlement--Fusion, SCWR and NGNP (Ni bearing alloys), Spallation Specific Issues of Common Interest (continued) • Very low dose rate, long time exposures • Wide range of dose rates < 10-10 to > 10-7 • High temperature deformation mechanisms—Fusion Demo/Power Plants, Gen IV • Low temperature flow localization and fracture • Behavior of structural composites (mechanical properties and structure)-- Fusion, NGNP, GFR • Liquid metals in contact with irradiated structural materials--Fusion, LFR, Spallation • Water coolant in contact with irradiated structural materials--ITER, SCWR, Spallation • Gas coolant in contact with irradiated structural materials--Fusion, NGNP, GFR Key Cross-cutting Phenomena in Theory and Modeling of Structural Materials

• hardening and nonhardening embrittlement including underlying microstructural causes and the effects of helium on fast fracture; low temperatures and all doses • flow localization, consequences and underlying microstructural causes, low to intermediate temperatures, all doses • high temperature deformation and fracture, including helium effects; higher temperatures and doses • irradiation creep and thermal creep; intermediate and high temperatures; intermediate to high doses • swelling and phase stability, intermediate temperatures; intermediate to high doses • welding, joining and processing issues • fatigue and creep-fatigue interactions; dependent on cyclic loading; load level and total cycles • hydrogen and interstitial impurity effects on deformation and fracture • chemical compatibility, erosion, bulk corrosion, stress corrosion cracking, impurity and corrosion product transport Radiation damage is inherently multiscale with interacting phenomena ranging from ps-decades and nm-m Comparison of Fission and Fusion Radioactivity after Shutdown

10-1

Fission: 10-3 Light Water Reactor Fusion: Conventional Ferritic steel

10-5 Fusion: Reduced Activation Ferritic Steel Coal Ash 10-7 Below Regulatory Concern Fusion: Silicon Carbide Composite 10-9

Curies/Watt (Thermal Power) Fusion: Vanadium Alloy

10-11 1 10 100 1000 104 Years After Shutdown