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Overview of the INL Fusion Safety Program and Summary of Work in Progress

IAEA First Technical Meeting on the Safety, Design and Technology of Fusion Power Plants

Vienna International Centre

Vienna, Austria

May 3, 2016

Brad Merrill, Lee Cadwallader, Masashi Shimada,

Chase Taylor, Robert Pawelko and Paul Humrickhouse

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• Give an overview of the Fusion Safety Program (FSP) at the Idaho

National Laboratory (INL) in the context of FSP’s role in national and

international collaborations

• Describe ongoing FSP experimental capabilities in tritium and

materials research and computer safety code development

• The FSP is a well established program in the US that has an important

role in supporting national and international licensing, safety, and

environmental assessments for magnetic fusion

Presentation Outline

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• INL is the US lead laboratory for Magnetic Fusion Energy (MFE) Safety

– Office of Fusion Energy Sciences (FES) awarded this contract in FY-1979

• FSP’s Mission: assist the fusion community in developing the inherent safety

and environmental potential of fusion power:

– Developing fusion licensing data and analysis tools

– Participating in national and international collaborations and design

studies

– Assist the US and international fusion community in licensing activities

and guidance in operational safety

• FSP maintains core competencies in the following areas:

– Tritium retention and permeation in fusion plasma facing component

(PFC) and blanket materials (also be applied to fission applications)

– Dust and radiological source characterization, dust collection, and material

oxidation

– Fusion safety code development

– Risk Assessment, codes and standards, and failure rate data collection

– Waste management strategies

INL Fusion Safety Program

DOE National and International Collaborations

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• US/Japan collaboration on tritium plasma, heat and neutron irradiation experiments (PHENIX)

– Studying tritium retention and permeation in irradiated tungsten for fusion chamber walls (2013-2019)

• US Fusion Nuclear Science Facility (FNSF) Project

– A collaboration to outline the mission for the first US nuclear fusion device, envisioned to bridge the gap from ITER to DEMO

• USIPO, F4E and ITER International Organization (IO)

– RAMI contract that includes training classes for US designers

– MELCOR support for F4E safety analyses through a contract

– ITER TBM PC, ITER IO magnet arcing safety contract

• US/Korea Fusion Collaboration

– UCLA and INL are collaborating with Korea’s NFRI on fundamental research and development for fusion nuclear science from 2014-2016.

• The International Energy Agency’s multilateral technology initiative on the Environmental, Safety and Economic Aspects of Fusion Power

– China, EU, Japan, South Korea, Russia and the US. Demonstrate the potential for Fusion Energy as a very safe, environmentally attractive and inexhaustible source of power (2012-2017).

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INL Fusion Safety Program (FSP) Overview

Tungsten clad divertor

Designed operating temperature up to 3000 C

• FSP’s Mission: assist the fusion community in developing the inherent safety and

environmental potential of fusion power:

– Developing fusion licensing data and analysis tools

– Participating in national and international

collaborations and design studies

– Assist the US and international fusion community in

licensing activities and guidance in operational safety

• INL FSP input to every ITER safety report including

its RPrS:

– Licensing guidance (DOE Fusion Safety Standard,

1996) & reviews

– Computer code development (modified MELCOR

code), system modelling & accident analysis

– Tritium in-vessel retention data, material oxidation

and dust characterization and explosion indices data

at FES’s Safety and Tritium Applied Research -

STAR facility

– Component failure rate data collection, RAMI

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• DOE FES’s tritium and safety material’s

data research facility since the early

1980’s, expanded in 2001 to become

STAR

• It’s a DOE less-than Hazard Category 3

Nuclear Facility capable of handling low

radiation levels (e.g., W, Ni, Mo, T2) :

contact dose < 1.5 mSv/h, annual dose

< 7 mSv/y, T2 inventory <1.5 g

STAR’s research priorities continue to be: 1. In-vessel tritium source determination:

− Tritium retention in plasma facing

components (PFCs)

2. Blanket & Ex-vessel tritium permeation

testing:

− Tritium permeation through PFCs

and blanket/structural components

3. In-vessel dust source term evaluation:

− Dust characterization and explosion

potential

Safety and Tritium Applied Research (STAR) Lab Located at the INL Site…

STAR

Remote operation center outside PermaCon

• Only linear plasma device in the world that can study tritium inventory and permeation neutron irradiated targets

– Neutron damage creates defects in materials that are different from high energy ion induced damage that leaves additional interstitial atoms in the metal. These defects trap tritium as it permeates through the material

– Tungsten irradiated to 3.0 dpa in ORNL’s High Flux Isotope Reactor (HFIR) will be studied in TPE under the US/Japan PHENIX collaboration

• TPE has been upgraded to produce more prototypical ITER divertor conditions: ion flux 1x1023 m-2s-1, heat flux ~4 MW/m2, and W temperature 1000ºC

Source side

plasma

Target side plasma

TPE TPE’s glovebox TPE’s PermaCon Structure

Tritium Plasma Experiment (TPE) at STAR

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Accelerator

Analysis line for NRA

Chamber for housing

samples and detector for

NRA

p transport calculation

System at the

University of

Wisconsin

W samples irradiated in HFIR exposed to TPE plasma

(M. Shimada, Physica Scripta T145 (2011) 014051).

Nuclear Reaction Analysis (NRA) of Tungsten Samples from TPE

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First of a kind data

Thermal Desorption Spectroscopy (TDS)

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• The Thermal Desorption Spectroscopy (TDS) includes 6 calibrated leaks (3 He and 3 D2) and is capable of distinguishing He (4.0026 amu) and HT (4.0239 amu) from D2 (4.0282 amu)

• Typical irradiated W (0.025 dpa) desorption spectrum obtained at 0.15 K/s

First of a kind data

22Na (0.5 MeV) or 68Ge (1.9 MeV)

• To predict trapped tritium inventory

computer codes need defect density

• PAS is the tool we are using to

determine this parameter

Positron Source Thermalize

Diffusion

γ-ray

γ-ray

0.511 MeV

0.511 MeV

Positron Annihilation Spectroscopy (PAS) System: a Method for Quantifying Defect Density

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S Parameter: Annihilation with low

momentum valence and unbound electrons,

open defects enhance these events

*C. N. Taylor, et al. “Development of positron annihilation

spectroscopy for characterizing neutron irradiated tungsten,”

Physica Scripta T159 (2014) 014055 (5pp)

• Initial comparison of NRA and PAS results*

• Because of the high trap energy for W (~1.6 eV),

NRA measurements for TPE low temperature

targets (< 200ºC) are more indicative of defect

density

Positron Annihilation Spectroscopy: Doppler Broadening

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First of a kind data

Key Safety Issues: chemical reactivity (explosions) and toxicity, radioactivity content

Comparison of Size Distributions CMD (µm) + GSD

Machine

Lower Regions Middle Regions Upper Regions

DIII-D 0.66 + 2.82 0.60 + 2.35 0.89 + 2.92

TFTR 0.88 + 2.63 1.60 + 2.33 -

Alcator-Cmod 1.58 + 2.80 1.53 + 2.80 1.22 + 2.03

JET 27 + (-) - -

TEXTOR 5-20 + (-) - -

Tore Supra 2.68 + 2.89 2.98 + 2.94 3.32 + 2.94

ASDEX-Upgrade 2.21 + 2.93 3.69 + 2.81 3.59 + 3.08

LHD 8.59 + 2.67 6.31 + 2.39 8.73 + 2.09

NOVA 1.12 + 1.90 0.76 + 2.03 0.90 + 1.93

Particle Size Distribution, Specific Surface Area, Surface Mass

Density, Composition, Shape, and Tritium Content

• Focus mostly on in-vessel dust

inventory from plasma erosion-

induced generation

• Dust characterization is an

important step in understanding

the safety concerns

• Beryllium dust explosion indices

have been measured for ITER

licensing

Vacuum collection technique

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Recent ASDEX Upgrade dust characterization on Si witness

plates: Balden, Humrickhouse, et al , Nuclear Fusion 54

(2014) 073010.

Dust Generation, Characterization, and Mobilization Used for ITER Licensing

Experimental Chamber for Evaluation of Exploding Dust

(ExCEED)

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• ExCEED took first of a kind data for ITER on beryllium dust explosions in FY13 but has been shut down since that time

• Further data on different particle sizes and dust/hydrogen mixtures are required to build a meaningful database

• These data are critical needs for ITER licensing

First of a kind data

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Tritium Gas Absorption Permeation (TGAP) experiment • The experimental apparatus is a tube furnace (<700C)

inside a ventilated enclosure

• Testing of tungsten and low activation steel with 0.1 to

100 ppm T and hydrogen of 10-1,000 Pa.

TGAP Test section for PHENIX 6mm W permeation • 6 mm OD tungsten disc

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• This test section for a 6-mm W disc was tested at >550 C

FSP Safety Code Development Used in ITER Safety Analyses • MELCOR modified for ITER - a fully integrated, engineering level thermal-hydraulics

computer code that models the progression of accidents in fission and now fusion

power plants, including a spectrum of accident phenomena such as reactor cooling

system and containment fluid flow, heat transfer, and aerosol transport

• MAGARC - a coupled electromagnetic, radiant energy transport and heat conduction

code developed to analyze magnet arcing accidents,

• TMAP - tritium migration analysis program; it treats multi-specie surface absorption

and diffusion in composite materials with dislocation traps, plus the movement of

these species from room to room by air flow within a given facility

Code Users

• MELCOR for fusion has more than 25 users in US, JA, KO and EU: ITER IO, JAERI,

KAERI, ENEA, CCFE, CIEMET, KIT, LLNL, NFRI, AMEC, Studsvik, University of

Pisa, Kyung Hee University and Ulsan National Institute of Science and Technology.

• TMAP – unknown but has been listed as the third most requested code from the

DOE’s Energy Science and Technology Software Center (ESTSC)

Code applications • MELCOR: ITER’s NSSR (1&2), GSSR, RPrS, US DCLL TBM, reactor design studies

APEX (Li and FLiBe), ARIES AT/SC/ACT, LIFE, and the JA and EU DEMOs.

• TMAP: Used in safety analyses for ITER, US DCLL TBM, reactor design studies

APEX (Li and FLiBe), ARIES AT/SC

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TMAP Modified to Treat Any Number of Traps • TMAP multiple trap model compared

with TPE TDS data for 0.025 dpa

tungsten exposed to TPE plasma at a

temperature of 200ºC

• Implantation flux of 5x1021/m2-s

applied was for 7200 s (2 h)

• Six separate traps were assumed

(energies of 0.9, 1.0, 1.1, 1.2, 1.4, &

1.6 eV) at a total trap density of 0.35

at%

• However, the TPE plasma had to be

ramped down at a slower rate than

what actually occurred for this

calculation to allow the 0.9 eV traps to

saturate (TPE plasma ramp down in

300 s instead of 20 s)

• An explanation for deuterium traps at

low energy (0.9 eV) could be surface

blisters

300 400 500 600 700 800 900 1000 1100

TDS Temperature (K)

TPE TDS

TMAP multiple traps

TMAP multiple traps (delayed

plasma termination)

0.0

2.0x1017

4.0x1017

6.0x1017

8.0x1017

1.0x1018

1.2x1018

1.4x1018

Deu

teri

um

Flu

x (

D/m

2-s

)

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MELCOR-TMAP Code Development

• A new version of MELCOR for Fusion is under development based on the

most recent F77 MELCOR 1.8.6 (released in 2009) that combines the

capabilities of MELCOR for ITER (version 1.8.2) and MELCOR Multiple

Fluids (version 1.8.5).

• In developing MELCOR 1.8.6 for Fusion, the multiple fluids modifications

were made to this version of the MELCOR code while leaving the standard

MELCOR water properties intact. This approach has the advantage of

combining both previous versions (e.g., 1.8.2 and 1.8.5) in to a single code.

• Another advantage this new version is that its predictions have double

precision accuracy, unlike the versions 1.8.2 and 1.8.5 for Fusion which have

only single precision accuracy.

• This version of the code (MELCOR 1.8.6 for Fusion) was made available to

the MELCOR user community through SNL-NM during August of 2015.

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MELCOR Code Development (cont.) • Finally, the TMAP code is being imported into this MELCOR 1.8.6 for fusion

code in order to produce a more self-consistent safety accident analysis.

• Merrill’s ISFNT-12 paper covered this code development effort and its

application to a DEMO relevant water-cooled tungsten divertor.

Surface heat and particle fluxes of 10 MW/m2 and

5 × 1023 ions/m2-s, W thickness of 8 mm, water

temp 150 C, six trap findings from TITAN R&D and

trap density associated with 0.3 dpa

*Wampler, Nuclear Fusion, 49, 11502 (2009)

Single trap test for un-irradiated tungsten.

A simple equation* predicts 220 h to

breakthrough.

Future Directions in Fusion Safety Code

Development

MELCOR-TMAP

• Molecular dynamics solution (MDS) calculations to inform cluster

formation and diffusion, micro-void formation and transmutation trap

formation as a function of helium density

• There are a few improvements that need to be completed before

MELCOR-TMAP 1.8.6 for Fusion will be released to the user group during

2016.

• The US Nuclear Regulatory Commission (NRC) requested all fusion

modifications imported into the latest F95 Fortran version of MELCOR 2.x

under development at SNL-NM.

TMAP

• The development effort for a multi-dimensional version of TMAP that relies

on advanced mathematical solution packages capable of executing on

parallel computers (INL’s Multi-Object Oriented Simulation Environment)

and eventually coupled to RELAP7 is being considered.

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Operating Experience Data Collection supports Safety and RAMI objectives

• Safety needs for failure rate data include evaluating in-vessel components, e.g., new joining techniques and new materials. Some additional needs are for new components specified in new designs.

• ITER RAMI is using the IEA failure rate database developed by the US/EU/J. Data came from tokamaks (DIII-D, JET, FTU) and fusion facilities (TSTA, TLK, TPL).

• Requests from ITER designers show that more RAMI data are needed for tokamak systems, including cryogenics, magnets, plasma heating, fueling, diagnostics, and other systems.

• In 2015, operations data from the LANL Tritium Systems Test Assembly were found in storage. We plan to analyze these data for failure rates, repair times, and for operations information.

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FSP participates in a fusion design code activity

• The American Society of Mechanical Engineers (ASME) initiated a working group (WG) on a design code for fusion energy devices.

• The WG thus far has developed a “roadmap” or plan for writing design rules, and has a table of contents developed for the fusion device subsection within ASME Section III.

• The effort is moving forward. The WG leader, Ken Sowder, is seeking help from fusion designers in nations outside the US.

• There has been interest from the ITER IO, the UK, Korea, Japan, China, and other countries.

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Waste Management

• The DOE fusion safety standard (DOE STD-6002-96) states that waste, and especially high level waste, shall be minimized

• Fusion generates a large volume of low level waste; this may be reduced if we can recycle material

• The technical approach and economic aspects of recycling have not been sufficiently investigated

– Feasibility of separations processes have not been demonstrated

– Avoidance of high level waste generation in the process needs to be verified

• We collaborating with the University of Wisconsin on this issue through the US Fusion Nuclear Science Facility (FNSF) design activity

Future Directions in FSE Research

• Strategy based on FES guidance and 2013 FES Peer Review Comments

– Materials Research: Fusion materials, including tungsten irradiated, will be studied at high temperature and heat flux to measure tritium retention and permeation. Dust explosion measurements for fusion materials will continue in support of licensing and computer code development activities. New material diagnostics.

– Code Development: for the near term, a newer version of MELCOR for ITER will to be completed that includes tritium transport and dust explosion models. Long- term: Multi-dimensional safety code capabilities needs to be developed that take advantage of parallel computing (example RELAP 7)

– Risk and Licensing: FSP’s evolving failure rate database will be expanded to include maintenance data from existing tokamaks. Risk-informed safety analysis methods (example RISMC Toolkit) will be studied for application to an FNSF. Continue ASME codes and standards and licensing framework development.

– Collaborations: Participation in existing collaborations to leverage other institution's capabilities and reduce duplication of effort. STAR will move towards being more effective FES User Facility.

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The National Nuclear Laboratory