Neutronics Issues to be Resolved in ITER Test Blanket Module (TBM)  Demonstration of tritium self-sufficiency for a particular FW/B/S concept  Verification of the adequacy of current transport codes and nuclear data bases in predicting key neutronics parameters  Verification of adequate radiation protection of machine components and personnel.

Testing Tritium Self-sufficiency in ITER  This does not appear to be possible in ITER (e.g. basic shielding blanket does not produce tritium, no interface with tritium processing system in the TBM’, partial coverage vs.. full coverage in DEMO, etc. )  Direct demonstration of tritium self- sufficiency requires a fully integrated reactor system, including the plasma and all reactor prototypic nuclear components.

Testing Tritium Self-sufficiency in ITER (cont’d)  We should rely on indirect demonstration by utilizing the information obtained from local and zonal tritium production rate in the TBM and the associated uncertainties in their prediction, in addition to information on tritium extraction and flow in TBM, and extrapolate this information to DEMO and power-producing reactor conditions. This seems to be a difficult task

Testing Tritium Self-sufficiency in ITER (cont’d)  We should rely on indirect demonstration by utilizing the information obtained from local and zonal tritium production rate in the TBM and the associated uncertainties in their prediction, in addition to information on tritium extraction and flow in TBM, and extrapolate this information to DEMO and power-producing reactor conditions. This seems to be a difficult task Impact of Partial Coverage of TBM on Local Tritium Production Rate, TPR (compared to full coverage) Ratio Ratio of local TBR from 6 Li (T 6 ) and 7 Li (T 7 ) in the poloidal direction at front breeding surface to corresponding values in full coverage case

Classification of Neutronics Tests (A) Dedicated Neutronics Test Aim at examining the present state-of-the-art neutron cross-section data, various methodologies implemented in transport codes, and system geometrical modeling as to the accuracy in predicting key neutronics parameters (TBM of “Look-Alike” type) (B) Supplementary Neutronics Tests Intended to be performed in TBM (or submodules of “Act-alike” type) used for non-neutronics tests (e.g. thermo-mechanics test, etc).

(A) Dedicated Neutronics Tests  Unlike the case of using engineering scaling to reproduce demo-relevant parameters in an “Act- alike” test module, dedicated neutronics tests require a “Look-alike” test module for a given blanket concept. This is due to the desire to quantify a realistic error bars associated with various neutronics parameters when predictions with various codes/nuclear data are compared to measured data. An obvious example is verifying the potential for satisfying tritium self-sufficiency conditions.

(A) Dedicated Neutronics Tests (con’d)  It is strongly desired to perform neutronics tests on as “cold” module as possible to minimize problems associated with elevated breeding temperature (e.g. tritium permeation).  It is therefore recommended to perform these tests as early as possible during the DD operation phase (year 4) or in low duties cycle DT operation phase (year 5&6)

(A) Dedicated Neutronics Tests (con’d)  Each blanket concept (e.g. SB or MS) should have its own tritium fuel cycle that is isolated from ITER basic machine. Tests for tritium production rates, tritium permeation, transport and isotope separation will only demonstrate the potential of each blanket concept to generate and control tritium flow. It is by no means meant to confirm tritium self- sufficiency.

(A) Dedicated Neutronics Tests (con’d)  Tests for tritium production and comparison to measured data will only quantitatively assign error bars (uncertainties) in TPR prediction that will be useful in demonstrating the feasibility of meeting tritium self-sufficiency in a Demo where blanket full coverage and closed tritium cycle (tritium burn up rate in the plasma) are realized

(A) Dedicated Neutronics Tests (con’d)  Fundamental questions that need to be answered (in ITER at least under reduced reactor parameters and low fluence) is how to reliably measure tritium production in SB, LM, or MS breeders. This needs to be investigated by diagnostics’ people as well as experts’ opinion in this field

(A) Dedicated Neutronics Tests (cont’d) Measurements to be performed In-Pile Measurements Neutron and gamma heating rates and profiles Local (and if possible zonal) tritium production rates and profile. Neutron and gamma spectra at various location

(A) Dedicated Neutronics Tests (cont’d) In-Pile Measurements (con’d) Multi foil activation measurements (e.g. 27 Al(n,2n), 58 Mi(n,p), 27 Al(n,a), and 197 Au(n,g), etc). These MFA measurements is used for neutron spectrum quantification at low, intermediate, and high energy. Out-of-pile measurements: Dose behind test module and at cryostat, neutron yield from plasma and source characterization (part of plasma diagnostics)

Dedicated Neutronics Tests and Fluence Requirements/conditions Fluence 1 W s/m 2 to 1 MW s/m 2 ~ 1 MW s/m2 <1 MW y/m 2 Any linear combination of wall load and operating time which has this rage of fluence is acceptable. e.g,. 400 s (1 ITER 2.5 x n/cm 2 s is adequate. Can be performed in year 4 (DD operation) or during low duty cycle operation in DT operation (year 5&6) In-module parameters Out-of-module parameters Local tritium production rate, neutron Plasma neutron source and Gamma spectrum, nuclear heating, characterization. reaction rates. Dose behind shield and at cryostat. Test Module Conditions (Material, Geometry, Test Module Size) Full module (Look-alike) is preferable to preserve as closely as possible Demo-relevant conditions. Measurements to be carried out at the inner most locations in the module to minimize influence of boundary conditions from ITER shielding blanket (SS/H2O). Measurements are generic in nature for all blanket concepts. Operating Scenario Pulsed operation or Steady-state Out-of-module parameters H, He, dpa rates, activation Higher fluence is required to accumulate reasonable limits. Dedicated test facility other than ITER (i.e. IFMIF) may be necessary

Fluence Requirements for Some Measuring Techniques (To be confirmed by Experimentalists’ Group) 1 mW s/m 2 1 W s/m 2 ~ 1 kW s/m2 1 MW y/m 2 Integral Parameters Neutron yield NE-213 fission chamber Multifoil Activation (MFA) Liquid scintillators Tritium production rate Lithium glass detectors Gas counters Mass spectroscopy Proportional counters Thermoluminescent dosimeter (TLD) Nuclear heating Gas filled counter TLD Calorimeter Nuclear Reaction rate Fission chamber Activation foil Mass spectrometry Neutron and Gamma spectrum NE-213 proton recoil MFA

(B) Supplementary Neutronics Tests Intended to be performed in TBM of “Act- alike” type) used for non-neutronics tests (e.g. thermo-mechanics test, etc). Objectives  Provide additional supporting information to the dedicated tests in quantifying the uncertainties in prediction that could be used as safety factors in Demo blanket design

(B) Supplementary Neutronics Tests (con’d)  Provide the source term (e.g. heat generation and tritium production rate) for other non-neutronics tests devoted to predictive behavior and engineering performance verification (e.g. tritium permeation and recovery tests, thermo- mechanics tests, after heat removal tests, etc)  These tests can be scheduled during the high duty DT operation phase (year 7-10) devoted to the integrated tests.

Geometrical Requirements  Key neutronics parameters to be tested (e.g. heating rates, tritium production rates) should be prototypical to those found in Demo reactors. “Look-alike” TBM is required.  Neutronics parameters are sensitive to the TBM size and locations were measurements are intended to be performed. These locations should be selected away from boundaries as possible.

Geometrical Requirements (con’d)  Parameters to be tested should not change much over a short distance inside the TBM. It is preferable to have flat values over large distances to eliminate uncertainties in defining measuring locations

Objective of the present Analysis Examining the distances over which the nuclear heating rate and tritium production rate are constant inside the US TBM based on the Dual Coolant Pebble Bed (DCPB) ceramic breeder blanket concept. Two configurations are considered in the TBM:  Act-alike configuration (left Config.)  Look-Alike Configuration (right Config.)

Arrangements The US TBM is placed inside ¼ of ITER port Japan TBM, based on DCPB concept is placed next to the US TBM and also occupies ¼ of the port US DCPB: Li 2 TiO 3, 75%Li-6, Be multiplier Japan DCPB: Li2SiO4, 90%Li-6, Be multiplier

The U.S. Test Blanket Module Act- Alike Config. Look- Alike Config.

Japan Test Blanket Module

Details of the Model at Port

Japan Test Blanket Module Two Dimensional R-  del Calculation performed with DORT 2-D Discrete Ordinate Neutron-gamma Transport Code

Japan Test Blanket Module Nuclear Heating in the FW of the U.S. TBM in the Toroidal Direction

Japan Test Blanket Module Nuclear Heating in the FW of the U.S. TBM in the Toroidal Direction  Nearly flat over a distance of ~10-14 cm (left Config.) and ~16-20 cm (right Config.)- Flatness decreases by depth  Heating rate measurements could be performed over these flat regions with no concern for error due to uncertainty in location definition

Japan Test Blanket Module Nuclear Heating Across the U.S. TBM in the Toroidal Direction at depth 42mm behind FW

Japan Test Blanket Module Nuclear Heating Across the U.S. TBM in the Toroidal Direction at depth 42mm behind FW  Heating rate in the breeder of the left. Config. is a factor of ~4 larger than in Be of the Rt. Config. and is flat over ~10 cm. It peaks near the vertical coolant panels.  Heating profile in beryllium is flat over the entire layer. This feature is applicable to other beryllium layers (not shown)

Japan Test Blanket Module Radial Nuclear Heating in the left Config. at a Toroidal Distance of 28 cm from the Frame

Japan Test Blanket Module Radial Nuclear Heating in the left Config. at a Toroidal Distance of 28 cm from the Frame  Nuclear heating rates in the breeder layers are the largest.  Nuclear heating in Beryllium is the lowest.  Heating rates in the structure contents of the FW and horizontal cooling panels (HCP) have intermediate values

Japan Test Blanket Module Toroidal Profile of Tritium Production Rate (TPR) in each Breeder Layer of the U.S. TBM

Japan Test Blanket Module Toroidal Profile of Tritium Production Rate (TPR) in each Breeder Layer of the U.S. TBM  Profiles of the TPR is nearly flat over a reasonable distance in the toroidal direction where measurements can be performed (10-16 cm in the left Config. and cm in the right Config.).  Steepness in the profiles near the ends of layers is due to presence of Be Layer. TPR is a factor at these locations

Japan Test Blanket Module Radial Tritium Production Rate at a Toroidal Distance of 28 cm from the Frame

Japan Test Blanket Module Radial Tritium Production Rate at a Toroidal Distance of 28 cm from the Frame  The TPR profiles in the radial direction are much steeper than in the toroidal direction  The distance over which TPR changes by ±5% from its lowest value is limited to ~1 cm.  To achieve high resolution, TPR measurements should be performed within this 1 cm range in the radial direction