Fusion, Material and Safety

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Fusion,Material and Safety

• RUSA May 5 6, 2009
• Sture Nordlinder

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Outline

• Background and introduction
• Major projects since last RUSA
• HIP development
• Support to licensing, safety analyses
• Corrosion issues
• Ongoing projects and prospects
• Fire scenarios in Tritium and Hot-cell building
• Proposal on corrosion issues in vacuum vessel at
  emptying and drying

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Matter of organizations

• In the end of 2008 we finalized the last 5.1b
  contract
• No secondment at EFDA or F4E, (The person who
  previously had secondment has been direct
  employed by F4E)
• Studsvik will focus on the support for ITER via
  F4E, and in the future also DEMO
• Studsvik was the organization which received the
  first Grant agreement with F4E.

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Work program of Fusion for Energy

• Present program for 2009 can be found on the
  website of F4E.
• The objectives of F4E are
• Provide EUs contribution to ITER
• Implement broader approach
• Prepare for the construction of demonstration
  fusion reactor (DEMO) but no activities in the
  2008 and 2009 program

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Major projects concluded during the last year

• HIP development
• Support to licensing, safety analyses
• Corrosion issues

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Hot Isostatic Pressing (HIP)

• A method to produce solid metal component from
  metal powder
• The advances that component with complex geometry
  can be manufactured (ex. shielding blanket with
  cooling tubes)

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Hot Isostatic Pressing (HIP)

• Purpose Demonstrate mechanical properties at the
  powder-plate joints after HIP with low oxygen
  powder.
• Three samples capsules were manufactured and
  delaminated.
• Each capsule contain about 5 liter of powder
• EFDA contract 06-1387, TW5-TVM-LOWHIP2

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Hot Isostatic Pressing (HIP)

• Test specifications of the capsules

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Hot Isostatic Pressing (HIP)

• Samples from the capsule were mechanically tested
  and chemically analyzed.
• Conclusion
• Thermal treatment gave better mechanical
  properties at powder-plate joints.
• The getter did not show any improvement
• No significant chemical differences (C, S, O and
  N)
• Further studies needed for optimization of
  temperature during pre-HIP treatment.

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Safety assessment, thermal hydraulic analyses -
background

• Thermal hydraulic analyses were performed during
  2008 for accident sequences identified by ITER.
  The objective was to perform independent analyses
  to verify that results, from similar analyses
  performed by ITER, are reasonable.
• The work is performed in support for the
  licensing of ITER, the RPrS (Raport Préliminaire
  de Sûreté).
• Three different accident scenarios has been
  studied.
• (Grant agreement F4E-2008-GRT-02-01 (ES-FS))

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1. Multiple First Wall pipe break with
simultaneous Divertor failure
Break in First Wall coolant loops (break area
0.02 m2)

• Main objective To analyze pressure development
  in plasma chamber
• Results Pressure increase to higher level for
  small break (1 tube) in the Divertor, compared to
  large Divertor break (10 tubes). The cause is a
  slow, but long release of water to the Plasma
  Chamber.
• Pressure increase for both cases are within
  safety limits.

Break in Divertor coolant loop
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2. Large Divertor and ex-vessel coolant pipe break

• Main objective Assess the releases of dust,
  tritium and ACP to the environment.
• Controlled releases (via Suppression Tank
  ventilation system, and via ventilation of TCWS)
  and un-controlled releases (leakage from TCWS and
  Gallery).
• Results Releases are within limits.

1. Ex-vessel break in Divertor coolant loop,
inside TCWS vault
2. In-vessel break in Divertor coolant loop
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3. Failure of two Vacuum Vessel coolant loops

• Main objective Assess temperature development
  in Plasma Facing Components.
• Results Temperature within safety limits. Small
  (within limits) releases of tritium and activated
  corrosion products (ACP) to the environment.

Break water is dis-charged into Gallery
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Recent work at Studsvik on corrosion issues

• Radiolysis modeling and estimations of corrosion
  potentials
• Experiments to study the effect of copper on the
  stress corrosion resistance of stainless steel
• Crevice corrosion tests of stainless steels in
  simulated Vacuum Vessel environment

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Radiolysis modeling and estimations of corrosion
potentials

• Radiolysis and corrosion potential estimates in
  ITER heat transfer systems
• In order to evaluate if metallic components will
  suffer from corrosion, it is of fundamental
  importance to know what corrosion potential the
  metal will acquire under the relevant exposure
  conditions.
• EFDA contract 07-1712, TW6-TVM-CUIMP

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Radiolysis decomposes the water

• The major source of oxidants in the cooling water
  is radiolysis, if no countermeasures are
  undertaken. This process splits the water
  molecule and as a result the oxidants H2O2 and O2
  are formed along with the reducing agent H2 and a
  number of radicals such as OH.
• Depending on the nature of this process, and the
  boundary conditions set by the initial water
  chemistry and radiation, the resulting corrosion
  potential can be either oxidizing or reducing.

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Modeling of corrosion potential (ECP)

• A model developed at Studsvik and validated for
  certain BWR and PWR conditions was adopted for
  ITER conditions
• The Studsvik model is thus able to estimate the
  corrosion potential as a function of oxidant
  content (H2O2 and O2), dissolved H2 content,
  temperature and flow rate.
• By using the input from radiolysis calculation,
  the model can be used to estimate the corrosion
  potential in ITER cooling system when plasma is
  burning as well as under idle condisions.

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Results of radiolysis and corrosion potential
modeling

• The calculations show that if no H2 is added,
  considerable amounts of H2O2 and O2 will be
  formed through radiolysis. The corresponding ECP
  will thus be oxidizing during plasma burning.
• On addition of H2 the level of radiolytically
  formed oxidants is suppressed, however, the
  conditions will most likely still be oxidizing
  according to the ECP model during plasma burning.
  
• Without the radiation from the plasma the
  oxidants will be consumed and less oxidizing or
  reducing conditions will be obtained. During the
  ITER pulse operation the ECP will show a cyclic
  behavior.

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Slow strain rate tests of stainless steel in
environment with Cu impurities

• In the primary heat transfer system of the ITER
  divertor both stainless steel and copper
  materials are used.
• In addition to the potential corrosion risk of
  especially the copper alloy an indirect effect of
  copper on the corrosion of stainless steel is
  considered.
• Due to corrosion of copper some material will
  dissolve and copper will be present in the
  cooling water.
• Copper is considered to be detrimental with
  respect to local corrosion of stainless steel.
• However, no data for ITER relevant materials in
  relevant environments are available.
• (Underlying technology/general support)

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Slow Strain Rate Test
During a slow strain rate test (SSRT) a tensile
specimen is slowly pulled to fracture (or to a
given elongation/stress) under simulated ITER
condition The fracture surfaces are studied after
the test to check if intergranular attacks
(IGSCC) have occurred.
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Test specifications

• Two different types of steel (304 and 316L(N) IG)
• Three levels of Cu-concentration (0, 10 and 100
  ppb)
• Test conditions
• Temp 200 degrees C
• Pressure 44 bar
• Strain rate was set to give fracture in two weeks

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Conclusions

• Stainless steel Type 316L(N) IG did not show
  stress corrosion cracking under the experimental
  conditions used.
• Sensitized stainless steel Type 304 showed high
  susceptibility to IGSCC under the same
  experimental conditions and also without Cu
  addition. Presence of 10 ppb Cu increases the
  susceptibility, whereas no significant difference
  could be observed between 10 and 100 ppb Cu.
• The presence of Cu in the water appears to
  decrease the electrochemical corrosion potential
  of both types of materials.

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Crevice corrosion test of Vacuum Vessel shielding
plates

• Crevice corrosion of the shielding blocks in the
  VV has been identified as a potential corrosion
  risk
• A preliminary test to check the crevice corrosion
  resistance of the different materials has been
  performed
• The test has been performed on five different
  types of stainless steel.
• ConclusionThe tests demonstrate that crevice
  corrosion is a concern both at during drying and
  wetting conditions.
• (Underlying technology/general support)

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Ongoing projects

• Studsvik has only one project running at the
  moment
• Detailed analyses of fire reference eventswith
  the purpose to study pressure and temperature
  transient at fire scenarios in tritium plant and
  hot cell, with the purpose to show amount of
  releases of toxic and radioactive product.
• Preliminary results show intensive fire resulting
  in releases above acceptable limits.
• The outcome will be discussed with F4E and ITER
  May 20.
• Grant agreement F4E-2008-GRT-01 part 2 (ES-SF)

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Future fusion activities at Studsvik

• Studsvik will focus on projects from F4E related
  to ITER but also DEMO
• One grant application is in the process at F4E
  regarding corrosion issues
• Experimental study providing corrosion data using
  nitrogen or hot dry steam to simulate drying
  sequence of the vacuum vessel cooling loops.
• Thermohydraulic modeling of the drying sequence
  of the vacuum vessel cooling loops.
• Experimental study on slow strain rate testing on
  joints of cooling pipes of Inconel 625, CuCrZr
  and 316L(N)-IG in presence of copper impurities,
  which is valid for the divertor
• Call for proposal F4E-2008-GRT-21 (MS-VV)

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