3 One- and Two- Dimentional Scoping Calculations

ITER G 73 DDD 2 01-06-06 W0.1
Nuclear Analysis Report
NAG-201-01-06-17-FDR

ITER G 73 DDD 2 01-06-06 W0.1
H. Iida, V. Khripunov, L. Petrizzi
Nuclear Analysis Group,
ITER Garching Joint Work Site
Nuclear Analysis Group, Joint Central Team and Home Team Contributors
M. Angelone (EU) P. Batistoni (EU) A. A. Borisov (RF)
H. Freiesleben (EU) U. Fischer (EU) Y. Chen (EU)
T. Inoue (JA) I. A. Kartashev (RF) H. Kawasaki (JA)
C. Konno (JA) F. Maekawa (JA) Y. Morimoto (JA)
K. Ochiai(JA) M. Pillon (EU) D. Richter (EU)
S. Sato (JA) G. Ruvutuso (EU) K. Seidel (EU)
O. L. Schipakin (RF) A. G. Serikov (RF) G. E. Shatalov (RF)
K. Shibata (JA) S. V. Sheludjakov (RF) S. Unholzer (EU)
T. Utsumi (JA) D. Valenza (JCT) F. Wasastjerna (EU)
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ITER G 73 DDD 2 01-06-06 W0.1
Contents
1 Introduction___________________________________________________________ 7
2 Basic Nuclear Parameters and Radiation Design Limits _______________________ 8
2.1 The Main Operating Parameters___________________________________________ 8
2.2 Reactor Component Performance Requirements _____________________________ 9
2.2.1 Static Heat Loads Specification for Magnet System ________________________________ 9
2.2.2 Radiation Limits to Magnets __________________________________________________ 9
2.3 Operating Scenarios (“M-DRG1”) ________________________________________ 10
2.4 Radiation Shielding and Criteria for Personnel Access________________________ 11
2.4.1 ALARA Target Threshold for Dose Rate________________________________________ 12
2.4.2 Radiation Access Zones and Conditions ________________________________________ 12
2.5 Critical Nuclear Responses in Structural Materials __________________________ 13
2.5.1 Rewelding Limits for Stainless Steel and Boron Content ___________________________ 14
2.5.1.1 He Production Limits_____________________________________________________ 14
2.5.1.2 Boron Content Limitation _________________________________________________ 14
2.5.2 Metallurgical and Radiological Limits of Impurities _______________________________ 15
2.5.2.1 Cobalt Specification______________________________________________________ 15
2.5.2.2 Niobium Specification____________________________________________________ 15
3 One- and Two- Dimentional Scoping Calculations ___________________________ 16
3.1 Calculation tools and calculation model____________________________________ 16
3.2 Radiation Fluxes and Nuclear Heat Distribution during Operation_____________ 19
3.2.1 Radiation fluxes ___________________________________________________________ 19
3.2.2 Nuclear heat distribution ____________________________________________________ 20
3.3 Helium production rate _________________________________________________ 24
3.4 Damage in the Blanket, Vacuum Vessel and Divertor Materials _______________ 25
3.4.1 Spectral Effects____________________________________________________________ 25
3.4.2 Spatial and Material Dependence______________________________________________ 26
3.4.3 Peaking Factors ___________________________________________________________ 28
3.4.4 Damage in the First Wall Fastener Materials _____________________________________ 28
3.4.5 Vacuum Vessel____________________________________________________________ 30
3.4.6 Damage Function of Neutron Fluence __________________________________________ 30
3.4.7 Displacements and other responses in the Divertor ________________________________ 30
3.5 Dose Rates during Operation and Shutdown _______________________________ 31
3.6 Decay Heat ___________________________________________________________ 34
3.7 Activation of air outside the Cryostat and Bioshield _________________________ 35
3.7.1 Location of the Assessment and calculated production rate__________________________ 36
3.7.2 Ar-41 concentration in the air_________________________________________________ 36
3.8 Water Coolant Irradiation_______________________________________________ 38
3.8.1 Critical Locations __________________________________________________________ 38
3.8.2 Nitrogen in the Blanket Water Coolant _________________________________________ 40
3.8.2.1 Residence Time _________________________________________________________ 40
3.8.2.2 16N and 17N in the Blanket Water Coolant ___________________________________ 41
3.8.3 Coolant Activation in Upper and Mid-Plane Diagnostic Ports _______________________ 42
3.8.4 Divertor Coolant___________________________________________________________ 43
3.8.5 Total 16 N Gamma-Ray Source in the Outlet Pipes inside the Cryostat_________________ 45
3.8.6 Absorbed Dose Rates and Coolant Pipe Activation________________________________ 45
4 Detailed Multi-Dimensional Analyses _____________________________________ 47
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4.1 Neutron Wall Loading Distribution on the First Wall_________________________ 47
4.1.1 Neutron Source Distributions_________________________________________________ 47
4.1.2 Averaged Values and Local Maximums ________________________________________ 50
4.1.3 Flux-to-Current Ratios ______________________________________________________ 51
4.2 Blanket_______________________________________________________________ 54
4.2.1 Heat Deposition in the Blanket Modules ________________________________________ 54
4.2.2 Flexible joint analysis_______________________________________________________ 56
4.2.2.1 Model description _____________________________________________________ 56
4.2.2.2 Radiation Damages of the Flexible Cassette (Ti alloy) and a Bolt (Incoloy) _______ 58
4.2.2.3 The Effect of Flexible Joint Module on Nuclear Heating in the TF Coil inboard Legs 59
4.2.3 Helium production in the branch pipe and at the surface of the vacuum vessel __________ 60
4.3 Vacuum Vessel ________________________________________________________ 63
4.3.1 Integral nuclear heat ________________________________________________________ 63
4.3.2 Local nuclear heat__________________________________________________________ 64
4.3.3 Gap streaming onto the vacuum vessel , ________________________________________ 70
4.3.3.1 Calculation model _____________________________________________________ 70
4.2.3.2 Results of Analysis ____________________________________________________ 71
4.2.3.2.1 Nuclear Heat on the Vacuum Vessel without Manifolds between Blanket Modules
(case 1) 71
4.3.3.2.2 Nuclear Heat on the Vacuum Vessel in case of V-shape Gaps with Manifolds between
Blanket Modules in the (case 2 and 3)_______________________________________________ 74
4.3.4 Heterogeneity Effects of the Vacuum Vessel on the Inboard TF coil Leg Nuclear Heating 76
4.3.4.1 Calculation Models ____________________________________________________ 77
4.3.4.2 Comparison of the Nuclear Heating Values for the Proposed Design and the ITER 3-D
Basic Model 79
4.3.4.3 Improvement of shielding efficiency by increasing water fraction (or decreasing steel
fraction) 82
Conclusion ______________________________________________________________________ 84
4.4 Divertor Cassette _______________________________________________________ 84
4.4.1 Modelling ________________________________________________________________ 84
4.4.2 Nuclear heat distribution, damage, Helium production and shielding capability _________ 86
5 Port Analysis_________________________________________________________ 92
5.1 ECH Upper port _______________________________________________________ 92
5.1.1 Nuclear heating in the magnet system __________________________________________ 93
1) Shutdown dose rates_____________________________________________________________ 93
5.2 NBI Port _____________________________________________________________ 95
5.2.1 Calculation model__________________________________________________________ 95
5.2.2 Results of the analysis ______________________________________________________ 98
5.2.2.1 Nuclear heating and insulator dose rate in the magnet system ___________________ 98
5.2.2.2 Neutron fluxes and dose rate after shutdown around the port ___________________ 99
5.3 ECH Ports ___________________________________________________________ 103
5.4 ICRF Port ___________________________________________________________ 107
5.5 LH Equatorial Port ___________________________________________________ 110
5.6 Test Blanket Modules in a Mid-Plane Port ________________________________ 112
5.7 Blanket Maintenance Port______________________________________________ 115
5.7.1 3-D analysis model and the method __________________________________________ 115
5.7.2 Results of the analysis _____________________________________________________ 116
5.7.2.1 Nuclear heating and insulator dose in the magnet system _____________________ 116
5.7.2.3 Dose rate as a function of time after shutdown______________________________ 122
5.7.2.4 Conclusion _________________________________________________________ 123
5.8 Radiation conditions inside the divertor ports _____________________________ 123
5.8.1 Introduction _____________________________________________________________ 123
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5.8.2 Model description_________________________________________________________ 124
5.8.3 Results for the RHP _______________________________________________________ 124
5.8.4 Results for the CPP________________________________________________________ 128
6 Radiation Properties of Diagnostic System Plugs___________________________ 134
6.1 Vertical Neutron Camera ______________________________________________ 134
6.1.1 Preliminary analysis and design consideration___________________________________ 134
6.1.2 Model __________________________________________________________________ 136
6.1.3 Nuclear performance ______________________________________________________ 136
6.1.3.1 Specific energy release __________________________________________________ 137
6.1.3.2 He-production in steel channel walls________________________________________ 137
6.1.3.3 Activation ____________________________________________________________ 137
6.1.4 Neutron and Photon Flux Distributions in the Camera Surrounding __________________ 137
6.1.4.1 “Background” Fluxes____________________________________________________ 137
6.1.4.2 Collimated Flux Components _____________________________________________ 138
6.1.4.3 Neutron spectra evolution along the channels_________________________________ 139
6.1.5 3-D Modelling ___________________________________________________________ 140
6.2 Edge Thomson Scattering System in Upper Port ___________________________ 141
6.3 LIDAR and Polarimetry Diagnostic Systems in the Integrated Mid-Plane Port __ 143
6.3.1 Plug model description_____________________________________________________ 143
6.3.2 LIDAR _________________________________________________________________ 144
6.3.3 Polarimetry System _______________________________________________________ 146
6.3.4 Other responses __________________________________________________________ 146
6.4 Motional Stark Effect Diagnostic System__________________________________ 146
6.4.1 3-D MSE Modelling_______________________________________________________ 148
6.4.2 Nuclear Response Distributions along the Diagnostic Channel______________________ 149
6.5 The In-Vessel Viewing System Study _____________________________________ 152
6.5.1 Streaming effects _________________________________________________________ 153
6.5.2 Shielding Block Efficiency__________________________________________________ 155
6.5.3 Findings ________________________________________________________________ 156
6.6 Photo-Neutrons in the Plasma Chamber __________________________________ 156
6.6.1 Incident neutrons and secondary photons_______________________________________ 157
6.6.2 Prompt photo-neutrons _____________________________________________________ 157
6.6.3 Impact on neutron diagnostics _______________________________________________ 158
6.6.4 Delayed photo-neutrons ____________________________________________________ 159
6.6.5 Effects of the delayed photo-neutrons _________________________________________ 161
7 Dose rate estimate outside bioshield ___________________________________ 162
7.1 Dose rate during reactor shutdown ______________________________________ 162
7.2 Dose rate during maintenance condition __________________________________ 163
7.2.1 Model description_________________________________________________________ 164
7.2.2 Results and discussion _____________________________________________________ 165
7.3 Dose rate outside the bioshield during operation ___________________________ 167
7.3.1 Calculation model_________________________________________________________ 167
7.3.2 Calculation results and discussion ____________________________________________ 168
8 The DD Phase Nuclear Performance ____________________________________ 172
8.1 Two component neutron source _________________________________________ 172
8.2 Neutron spectra and flux distributions____________________________________ 172
8.3 Neutron wall loading and fluence ________________________________________ 174
8.4 Nuclear energy deposition and power balance _____________________________ 175
8.5 D-D phase activation __________________________________________________ 176
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8.6 Concluding remark ___________________________________________________ 177
9 Uncertainty Estimate _________________________________________________ 178
9.1 Summary of T16, T218, T362 Experiments ________________________________ 178
9.2 T426 Experiment at FNG ______________________________________________ 179
9.2.1 Experiment set up_________________________________________________________ 179
9.2.2 Measurements____________________________________________________________ 180
9.2.3 Experiment analysis and results ______________________________________________ 181
9.2.4 Conclusions _____________________________________________________________ 189
9.3 T426 Experiment at FNS/JAERI ________________________________________ 190
9.3.1 Experiment set up_________________________________________________________ 190
9.3.2 Measurement ____________________________________________________________ 191
9.3.3 Experiment analysis and results _____________________________________________ 191
9.3.4 Conclusions _____________________________________________________________ 195
9.3.5 Summary _______________________________________________________________ 195
9.4 Discussion ___________________________________________________________ 195
9.4.1 Uncertainty in nuclear heating and helium production estimate _____________________ 195
9.4.2 Uncertainty in shutdown dose rate estimate_____________________________________ 197
10 Summary of Major Nuclear Responses and Conclusions ___________________ 199
10.1 Summary of Major Nuclear Responses ___________________________________ 199
10.1.1 Nuclear Heating in Superconducting Magnet System _____________________________ 199
10.1.2 Shutdown Dose Rates outside the Ports ________________________________________ 201
10.2 Conclusions __________________________________________________________ 203
Appendix A Chemical Compositions of Materials used for Nuclear and Occupation
Safety Analysis___________________________________________________________ 205
Appendix B Methodologies for dose rate calculation __________________________ 212
Appendix C ITER FEAT MCNP Models____________________________________ 214
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ITER G 73 DDD 2 01-06-06 W0.1
1 Introduction
The nuclear design requirements for ITER are determined from the operational phases that
are envisioned. The entire operation phase (integral first wall neutron fluence ~0.3 MWa/m 2 )
will last about twenty years and will involve a few thousand hours of D-T operation with the
tritium supplied from external sources.
Radiation transport calculations for prediction and confirmation of expected neutronic
parameters are an essential part of the reactor design process. Development and optimisation
of the design of blanket , vacuum vessel and other reactor components must be carried out in
a logical progression based on initial results obtained with one-and two- dimensional scoping
and parametric analyses followed by detailed three- dimensional radiation transport
calculations. The latter better characterise the radiation environment and accounts for
radiation streaming through ports, diagnostic systems, other penetrations, and the overall
geometric complexity of the tokamak system. For meaningful and self-consistent nuclear
analyses to be done, quality assurance and even improvement of the calculational tools and
nuclear data are essential.
The blanket is a water-cooled, stainless steel assembly that absorbs the heat lost by the
plasma through radiation and charged particles and by thermonuclear neutrons and secondary
gamma-rays produced in nuclear reactions and protects the vacuum vessel from excessive
neutron irradiation. The vacuum vessel is also a water-cooled, stainless-steel assembly with
borated- and ferritic-steel fillers, and provides, together with the blanket, radiation protection
for the superconducting magnets and other cryogenic components inside the cryostat.
The nuclear performance of ITER depends on the reactor operating conditions, radiation
limits specified for components and dose rate limits necessary to license the reactor. The
complex nature of a tokamak compels the nuclear analysts to accurately map the radiation
fields around the reactor. Component design and disposition of shielding must be performed
in collaboration with component designers to minimise activation to assure personnel safety
during operation and after reactor shutdown when access may be required to support
maintenance activities.
Mapping of the radiation distributions around the tokamak machine and inside the cryostat is
strongly dependent on the availability of up-to-date design. The results reported here are for
different phases of the ITER design that have evolved considerably over the course of the
EDA and its extension.
The complex nature of the calculational model requirements for three-dimensional radiation
transport analyses precludes rapid changes in the model descriptions. The results presented in
this document are the best updated one available at the end of EDA.
Conclusive nuclear design calculations have to be performed when "final" engineering
drawings become available. These studies and analyses must, by necessity, be completed in
ensuing ITER design and pre-construction phases. As a result of comprehensive nuclear
analyses with corresponding design measures, the nuclear responses and radiation conditions
both inside and outside the reactor meet the main DRG1 requirements.
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ITER G 73 DDD 2 01-06-06 W0.1
2 Basic Nuclear Parameters and Radiation Design Limits
The nuclear performance and the shielding efficiency of the ITER blanket, vacuum vessel,
ports and other penetrations in these assemblies are determined from specified radiation
constraints on the in-vessel and out-vessel components. Radiation shield and material
compositions and configurations are also determined by requirements for maintenance
including hands-on activities and remote maintenance operations.
Initially, the basic set of machine parameters and radiation limits for different components
and systems were chosen in reference 1 as a starting point for the preliminary ITER radiation
shield evaluations. In general, these were ultimately specified by component designers. The
physics and trade studies were proceeded in parallel with conceptual and engineering design
of the device to assure feasibility. These parameters were expected to change and expanded
as the evaluations and studies proceeded and as they provide feedback into the iterative
design process.
The basic parameters of the ITER device derived from both physics, engineering and
technology considerations and the main requirements for reactor components used in this
work were taken directly from the Design Requirements and Guidelines Level 1 (DRG1) 2
and other supported documents. These serve as the formal basis for the radiation shield
design development and for the following assessing nuclear performance of the ITER design.
A set of critical parameters arise from these design requirements that must be simultaneously
controlled to minimise radiation damage and nuclear heating in materials and structures, such
as:
damage and gas production rates in in-vessel components,
local nuclear responses in supeconducting magnets,
integral nuclear heating in the TFC and intercoil structures,
neutron fluxes and residual radiation doses at maintenance locations,
radiation conditions behind the cryostat, and
performance of diagnostic equipment.
2.1 The Main Operating Parameters
The next nominal (reference) operating parameters important for nuclear analysis are
specified in the DRG1 2 .
1 ITER-FEAT Outline Design Report (ODR). G A0 RI 99-11-22 W 0.1 November 1999.
2 Design Requirements and Guidelines Level 1 (DRG1). G A0 GDRD 2 00-12-01 W 0.5. December 1, 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 2.1-1 Main Operating Parameters
Fusion power 500 MW
Total average neutron fluence at the first wall 0.3 MWa/m 2
Integrated full power operation time 4600 h
Peak burn duty cycle 25 %
Nominal number of pulses 30000
The burn duration for the reference design is 400 s. However, 3000 s burn time and 12000 s
minimum repetition time are foreseen in the assessed non-inductive operation scenario II.
Besides, the machine design has to be assessed for the enhanced power operation at 700 MW,
and also for the integrated full power operation time 7600 h and total averaged neutron
fluence at the first wall of 0.5 MWa/m 2 .
2.2 Reactor Component Performance Requirements
2.2.1 Static Heat Loads Specification for Magnet System
According to the DRG1 specifications of heat loads to structures cooled at cryogenic
temperature (Table 2.2-1) are considered as the nominal value for the heat removal systems
(e.g. the cryoplant) and as a maximum allowable value for the thermal and nuclear shielding
systems.
Table 2.2-1 Heat Loads Specification for Magnet System
Total nuclear heating to TF Coils (maximum operating condition) 13.7 kW
total nuclear heating to TF Inner Legs 10.7 kW
Radiated power to magnet and cold structures from Thermal Shields:
normal conditions 2.7 kW
baking conditions 3.7 kW
Integral nuclear heating in the poloidal field (PF) coils using NbTi superconductor and in the
Nb3Sn central solenoid (CS) are not defined (restricted) exactly in the DRG1.
2.2.2 Radiation Limits to Magnets
One of the main functions of the vessel and the in-vessel components together shall provide
sufficient nuclear shielding to protect the superconducting coils.
Guidelines for assessing the structural design and requirement relating radiation effects for
the ITER magnets and their supports structures operating in the range 4-77 K are presented in
reference 1 .
In accordance with this document, it is assumed, at present, that “neutron fluences up to
5x10 22 n/m 2 do not cause any change of structural stability in any of the metallic material
properties except for copper.”
1 ITER Structural Design Criteria for Magnet Components. 21 June, 2001. N11 FDR 11 01-06-11 w0.1
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ITER G 73 DDD 2 01-06-06 W0.1
In addition to the total TFC heating the local nuclear heating and other superconducting
magnet design limits predetermining radiation shield design are specified in the DRG1.
These are specific nuclear energy release in the TFC case and the superconductor and the
peak dose arising from gamma radiation on organic insulation in any of the coils (Table 2.2-
2).
Table 2.2-2 Radiation Limits to Magnets defined in the DRG1
Maximum local nuclear heating in the cable conductor 1 kW/m 3
Maximum Local nuclear heat heating in the case and structures 2 kW/m 3
Peak gamma- and neutron- radiation dose to coil insulator *) 10 MGy
Total neutron flux to coil insulator 5x10 21 n/m 2
Notes :
*) This is a commonly accepted dose limit for epoxies used in ITER. Polyimides and
bismaleimides are more radiation resistant with experimental data showing only a
small degradation in shear strength at dose levels in excess of 100 MGy (See
reference 1 ).
Nuclear radiation on insulating systems has two impacts on the mechanical performance: (i)
break up of the long polymers (typically epoxy resin), and (ii) gas generation within the body
of the material, particularly hydrogen, oxygen and carbon dioxide. Thus, the allowable
radiation limits on the insulation have two components, both of which must be satisfied:
- Epoxy insulator must withstand a combined neutron and gamma-ray radiation
exposure of 1x10 9 rads (10 MGy) over the reactor lifetime. This includes the local gamma
dose arising from neutrons.
- The peak fast neutron fluence (>0.1 MeV) is 5x10 21 n/m 2 .
Both limit values, based on the available database, do not include safety factors for nuclear
radiation calculations.
As was mentioned before, a maximum assessed first wall neutron fluence of 0.5 MWa/m 2 is
assumed for the permanent reactor systems, such as the superconducting magnets and
vacuum vessel. It corresponds to about 0.9 year of the full power operation. Thus the
maximum allowable fast neutron flux limit in the insulator is ~1.8x10 10 cm -2 s -1 .
2.3 Operating Scenarios (“M-DRG1”)
In addition to the ITER fluence goals as described in Table 2.1-1, details of the fluence
scenario during the first ten years are given in the DRG1 as shown in Table 2.3-1. Based on
these specifications, the operating scenario “M-DRG1” was derived in reference 2 for
radiation hazard levels for occupation safety assessment (Table 2.3-2).
1 ITER Concept Definition, Vol. 2, ITER Documentation Series, No. 3. IAEA, Vienna, 1989.
2 K. Moshonas, Occupation Safety Assessment Specifications, Revision 2. G 82 MD 1 00-12-05 W 0.1.
Garching, December, 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 2.3-1 Neutron Fluence during the First Ten Years of Operation, MWa/m 2
Years: 1 - 3 4 5 6 7 8 9 10 Total
Equivalent number
of nominal pulses
0 1 750 1000 1500 2500 3000 3000 11751
Average neutron fluence
at FW, MWa/m 2 0 0 0.006 0.008 0.012 0.020 0.024 0.024 ~0.09
The “M-DRG1” assumes a fluence of ~ 0.09 MWa/m 2 for the first 10 years followed by
another 10 years of operation, this time with a fluence of ~ 0.21 MWa/m 2 .
However, two 6-day campaigns at 25% duty cycle are specified in “M-DRG1” at the end of
the first and second decades of ITER operation (Table 2.3-2, 2.3-3).
Table 2.3-2 M-DRG1 Specification and Input Assumptions (See reference Error!
Bookmark not defined.)
Average neutron first wall loading 0.56 MW/m 2
Total first wall neutron fluence for 20 years 0.3 MWa/m 2
Maximum duty cycle 0.25
Aggressive experimental campaigns 2 x 6 days
Total ITER operation time 20 years + 2 x 6 days
Fluence for the first 10 years + 6 days 0.094 MWa/m 2
Fraction of total fluence ~ 0.3
Fluence for the last 10 years + 6 days 0.21 MWa/m 2
Fraction of total fluence ~ 0.7
As it is mentioned in the DRG1, the outboard shielding blanket may be replaced with tritium
breeding blanket modules during the last 10 years of the ITER operation. However, for the
purposes of the nuclear performance and the SS/H2O-bulk shielding efficiency analyses the
“M-DRG1” scenario is used below in the report.
Table 2.3-3 “M-DRG1” Average First Wall Neutron Fluence
Years: 1 - 4 5 6 7 8 9 10 Total
Fluence, MWa/m 2
0 0.006 0.008 0.012 0.020 0.024 0.022 + 0.002 *) 0.094
Years: 11 - 20 Total
Fluence, MWa/m 2
0.208 + 0.002 *) 0.210
2.4 Radiation Shielding and Criteria for Personnel Access
Personnel access in the pit and behind the biological shield will be prohibited during reactor
operation. However, worker access at the TFC, and inside the cryostat may be necessary
after shutdown. In particularly, the coil terminals (connections of the coils to their
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ITER G 73 DDD 2 01-06-06 W0.1
superconducting busbars) located within the cryostat, near each individual coil and the
electrical insulating breaks in the cooling lines shall be accessible for disconnection and
reconnection, and for repair or replacement by means of hands-on operation.
Thus the bulk radiation shield including the steel/water blanket, the vacuum vessel, and other
in-vessel and out-vessel components together, shall provide sufficient nuclear shielding not
only to protect the superconducting coils, but also to reduce activation and residual dose rates
inside the cryostat at port areas. This reduction of the residual dose rate should be as low as
reasonably achievable to facilitate hands-on maintenance and emergency hands-on repair
operation.
In reference 1 , it was reported that one or two orders of magnitude additional neutron
attenuation is necessary to minimise shut-down dose rates in the cryostat compared to the
neutron attenuation necessary to minimise the TF coil heating.
2.4.1 ALARA Target Threshold for Dose Rate
The ALARA target threshold for dose rate 2 is less than 100 µSv/hour 10 6 s (~12 days when
the main residual activity from steel components has decreased by more than one order of
magnitude) after shutdown during and at the end of the DT operation period.
2.4.2 Radiation Access Zones and Conditions
According to the DRG1 the radiation shielding shall be designed to minimise the number of
components that require remote maintenance and also to permit personnel access in the
annular space outside the bioshield.
Shielding cells will be built around dedicated ports to allow parallel hands on maintenance in
adjacent volume when an activated components is in the cell. That is why the level of
ionising radiation outside the biological shield (with the exception of the NB cell)
surrounding the tokamak shall be limited to 10 µSv/hour 24 hours after shutdown, to allow
radiation workers unlimited access to those areas.
The accessibility of all areas of the ITER plant and personnel access limitations are defined in
reference 3 depending on the anticipated radiological hazard and conditions during
maintenance (See Table 2.4-1).
1 R.T. Santoro, H. Iida, V. Khripunov, M. Sawan, T. Inoue, ITER Radiation Shielding and Neutronics Analysis.
Fusion Engineering and Design, 39-40 (1998) 593-599.
2 Safety and Environmental Criteria. In: Plant Design Specification. G A0 RI 2 99-12-12 W 0.3.
3 Radiation Access Zones. In: Generic Site Safety Report, Volume VI, Occupational Safety. November, 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 2.4-1 Area Classification and Radiation Access Conditions, [2.13]
Access Limitation Total
(internal & external)
Dose Rate
A Free (unlimited) access
for all site personnel
B Supervised areas. Allowing
limited access for nonradiation
workers and
unlimited one for radiation
workers.
C Controlled and limited access
areas for all workers.
Appropriate radiation
protection and exposure
planning.
D Controlled/Restricted areas,
entry by exception
with a high level of approval.
Area Contamination
Characteristics
< 0.5 μSv/h No surface, airborne and
cross-contamination
< 10 μSv/h No loose contamination
tolerated
< 1 mSv/h Identified and controlled
contamination levels
maintained by ALARA
> 1 mSv/h,
exceeding those
allowable in Zone C
Permanent contamination
levels (or exceeding those
allowable in Zone C
Areas with limited access requirements dedicated for specific maintenance, such as the NB
cell and the areas inside the bioshield of the port maintenance areas, shall meet the
requirements for Access Zone C, 10 6 seconds after shutdown, and should be limited to 100
μSv/h, the ALARA guideline for allowing radiation workers hands-on access.
Areas where the guideline of 100 μSv/h is not met shall be reviewed for acceptability on an
individual basis.
2.5 Critical Nuclear Responses in Structural Materials
Rationale for materials selection for in-vessel components and magnets, including
permissible variation of main alloying elements and impurities, are given in several
documents (See, e.g. reference 1 ).
The detailed information on chemical composition of materials used for neutronic and
activation analysis is specifically given in reference 2 (See also Appendix A). All element
number densities in materials as based alloying elements as anticipated specified impurities
provided by the material manufacturer are in the range of permissible variations or
recommended values. A group of elements are elements with the specification of content
limitation due to ITER specific requirements. These are, for example, Boron limitation to
provide re-weldability of stainless steel, Cobalt and Niobium limitations in 316 type SS to
satisfy safety requirements.
1 G. Kalinin and V. Barabash, Material Assessment Report. G 74 MA 10 00-11-10 W 0.2. July 2001.
2 G. Kalinin and V. Barabash, Chemical Composition of Materials for ITER Components. G 73 MD 40 00-10-
20 W 0.2. 20 October, 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
2.5.1 Rewelding Limits for Stainless Steel and Boron Content
Existing ITER designs consider the possibility of the replacement of the divertor, and failed
components. It will be possible only if field welds are protected by sufficient shielding to
allow rewelding.
The design must also assure that all field welds will be reweldable after the same fluence.
That is why the effects of gaps between modules, divertor cassettes, and radiation leakage
through ports will be minimised by design.
The shielding blanket will be designed to maintain vacuum vessel reweldability and to
guarantee cutting and re-welding of the manifolds and branch pipe connections in case of a
major failure for the entire machine lifetime, until an average fluence of 0.3 MWa/m 2 is
reached on the first wall.
Besides, the replacement of the shielding blanket with a breeding blanket is considered in the
DRG1 at the end of the first ten yours operational period (at fluence 0.1 MWa/m 2 ).
In addition to these scheduled operations the design of the permanent components of the
machine such as the vacuum vessel and the toroidal field coils should be assessed to achieve
higher equivalent fluence levels up to 0.5 MWa/m 2 . According to the DRG1 all field welds
to vessel and in-vessel RH class 3 components shall be reweldable up to this fluence.
2.5.1.1 He Production Limits
Weldability of irradiated stainless steel is determined mainly by the helium generation. The
reweldability limits, provided by the allowable levels for the He production at weld locations
are the next (DRG1):
< 1 appm for thick plate welding, and
< 3 appm for thin plate or tube welding.
2.5.1.2 Boron Content Limitation
It is shown earlier (e.g., in reference 1 ) that the boron content in stainless steel has a
significant effect on the He generation in steel. In turn, the effect is most significant in steel
located close to water pipes due to the moderation of neutrons by water.
That is why, to achieve low He generation in SS, it is recommended that the boron content in
the steel be minimised.
0.002 wt.% B (or 20 ppm) in steel for the first wall, and
0.001 wt.% B (or 10 ppm) in steel for the vacuum vessel cooling tubes.
Helium production in the borated steel 304B7 used in the vacuum vessel as a filler shielding
material is not a constrain. 2 wt% B is specified for this non-structural material 12 .
1 R.T. Santoro, V. Khripunov, H. Iida et al., ITER Nuclear Analysis Report. G 73 DDD 1 98-06-17 W0.2
(NAG-101-98-06-17-CDR), Garching, June 1998.
Nuclear Analysis Report Page 14

ITER G 73 DDD 2 01-06-06 W0.1
2.5.2 Metallurgical and Radiological Limits of Impurities
The assessment of radiological hazard of stainless steel alloying elements and impurities
content (See reference 1 ) showed that the activation level of the isotopes, most dominating for
safety considerations, are 54 Mn, 56 Mn, 55 Fe, 57 Co, 58 Co, 60 Co, 57 Ni, 51 Cr and Nb 94 that
come from neutron reactions with the elements in the initial SS composition (Fe, Ni, Mn, Cr,
Co, and Nb). All elements, except Co and Nb, are main alloying elements and cannot be
markedly changed without having an impact on the stainless-steel properties.
The elements Co and Nb are relevant isotopes with regard to clearance and after all to a
potential reduction of waste masses of the vacuum vessel.
2.5.2.1 Cobalt Specification
Effect of Cobalt in irradiated stainless-steel is realised in the increased decay heat and
residual dose rate. Thus activated cobalt plays an important role for both occupational dose
and severe accidents like loss-of-coolant because Cobalt is a main component of the activated
corrosion products in the water cooling system.
Effects of Co reduction and the upper technical and economical reasons to limit Co content
have been assessed earlier 2 . It was recommended in reference 3 to adopt a 0.05 wt% Co
specification for the in-vessel steel SS 316 L(N)-IG [2.15].
The Co content is limited by the same value in SS 316 LN (or SS EC1) for superconductor
jackets, magnet structures, poloidal field coils, feeders and water cooling pipe lines 4 .
2.5.2.2 Niobium Specification
Nb is usually considered as an alloying element for stabilising austenitic structure and
preventing susceptibility to intergranular corrosion.
At the same time, Nb is an important element with regard to clearance. The activity
distribution implies that a significant fraction of the vacuum vessel has a potential for
clearance. The Niobium relevance is determined by its initial concentration in the reference
steel together with the long half live of the daughter radioisotopes, and, after all, by a
potential reduction of waste masses of the vacuum vessel (by separating the front part of the
outboard vacuum vessel).
In the Procurement Specification the niobium content in SS 316 L(N)-IG is limited to the
next values (See [2.14]:
- 0.1 wt % Nb in steel for in-vessel components, with exception of the vacuum vessel, and
- 0.01 wt % Nb for the vacuum vessel only, for waste disposal reasons, that will have
negligible impact on the cost.
1 General Site Safety Report, Vol. V. December 2000.
2 R.T. Santoro, H. Iida, V. Khripunov, The Effects of Impurities on the Neutronic Performance of ITER Grade
Stainless-Steel Type 316LN-IG. NAG-5-10-21-96.
3 R.T. Santoro, Material Compositions for the Blanket, Vacuum Vessel, Cryostat and Biological Shield. NAG-
38, 1997.
4 Material Specifications, in: ITER Magnet System Procurement Package 11.P2.
Nuclear Analysis Report Page 15

ITER G 73 DDD 2 01-06-06 W0.1
3 One- and Two- Dimentional Scoping Calculations
One-dimensional radiation transport calculations with discrete ordinate (Sn) code were
initially performed to guide and optimise the blanket and vacuum vessel design. In limited
cases (see 3.4) two- dimensional calculation was also conducted with the same purpose.
Detailed three-dimensional radiation transport calculations that more fully characterise the
radiation environment, account for radiation streaming through ports and other penetrations
in the blanket and vacuum vessel, and account for the geometric complexity of the tokamak
system were performed to assess specific problem areas. In this chapter, Sn code calculations
to characterise the nuclear performance of the blanket, vacuum vessel and bio-shield are
summarised 1 . Detailed 3-D analyses were subsequently fulfilled to determine specific nuclear
responses and related data in each ITER components (see Chapter 4,5 and 6).
3.1 Calculation tools and calculation model
The calculations were conducted with the discrete ordinate code ANISN 2 , the activation code
ACT-4 3 and the following nuclear data libraries;
1) Neutron and Gamma-Ray transport cross section FENDL2/175-42 groups 4
2) Kerma factors FENDL2 (with TRANSX 5 )
3) He production cross section FENDL2 (with TRANSX)
4) Displacement per Atom cross section FENDL2 (with TRANSX)
5) Activation cross section FENDL/A-2 6
6) Decay gamma-ray transport cross section THIDA-2 library/54 groups
Neutron cross sections from 1) to 4) are created assuming infinite dilution from MATXS files
which are based on the original FENDLE2 data. Activation cross section data 5) is the one
IAEA prepared for public use.
1 H. Iida, NAG-139 "Nuclear Response of ITER-FEAT estimated by 1-D calculation"
2 W. W. Engle, Jr., "ANISN, A One- Dimensional Discrete Ordinate Transport Code with Anisotropic
Scattering," K-1693 (March 1967), CCC-82, RSIC Computer Code Collection.
3 Y. Seki, H. Iida, H. Kawasaki, K. Yamada, "THIDA-2: An Advanced Code System for Transmutation,
Activation, Decay Heat and Dose Rate", Japan Atomic Energy Research Institute, JAERI 1301, March 1986.
4 A. B. Pashshenko, "Completion of FENDL-1 and Start of FENDL-2", IAEA Report INDC (NDS) - 352, 1996.
5 R. E. MacFarlane, "TRANSX 2: A Code for Interfacing MATXS Cross Section Libraries to Nuclear Transport
Codes," Los Alamos Laboratory Report, LA-12312-MS (July 1992).
6 A. B. Pashshenko, H. Wienke, J. Kopecky, J.-Ch. Sublet and R. A. Forrest, "FENDL/A-2.0 Neutron
Activation Cross Section Data Library for Fusion Applications," Version 1 of March 1996. IAEA-NDS-173,
March 1997.
Nuclear Analysis Report Page 16

ITER G 73 DDD 2 01-06-06 W0.1
Z
TF Coil V.V. Blanket
Inboard Plasma Outboard
R
TF coil or Inter-coil Structure
Cryostat Bioshieldt
Figure 3.1-1 One-dimensional calculation model of ITER
The model shown in Figure 3.1-1 is a "1D toroidal model" or " 1D annulus model", which
has inboard components, plasma and outboard components in all annular configuration. Two
“1D toroidal model” are used depending on the calculation purpose. When TF coil nuclear
response was required (case-TFC), outboard TF coil leg was placed outside the outboard
vacuum vessel. The TF coil leg was replaced with an inter-coil structure, when the dose rate
after shutdown was calculated (case-Dose). In the latter case cryostat and biological shield
were added as shown also in Figure 3.1-1. The dimensions of each component (layer) in the
“case-Dose", are given in Table 3.1-1.
Nuclear Analysis Report Page 17

ITER G 73 DDD 2 01-06-06 W0.1
Table 3.1-1. Dimensions of ITER components at equatorial plane (“case-Dose”)
Inboard Outboard
Zo Material Inner Outer mesh Zo Material Inner Outer mesh
ne
radius radius es ne
radius radius es
1 void 0 119.7 3 29 plasma 619 819 4
2 c-solenoid 119.7 216.9 30 30 scrape-off 819 842.1 1
3 void 216.9 219.9 1 31 be 842.1 843.1 1
4 TFC-case 219.9 242.4 8 32 f/w-cu 843.1 845.3 3
5 r-epoxy 242.4 243.6 1 33 f/w-ss2 845.3 847.6 2
6 cond1 243.6 293.6 16 34 f/w
(ss:65,h2o:31.5)
847.6 850.2 3
7 cond2 293.6 298.8 5 35 void 850.2 851.1 1
8 r-epoxy 298.8 300 1 36 blktwall 851.1 852.9 2
9 Tfc-case 300 307.5 7 37 Blkt
(ss:80,h2o:18)
852.9 855.3 5
10 void 307.5 318 1 38 Blkt
(ss:80,h2o:18)
855.3 862 5
11 t-shield 318 323.5 2 39 Blkt
(ss:88,h2o:10.8)
862 878.8 10
12 vv-case 323.5 329.5 6 40 Blkt
(ss:88,h2o:10.8)
878.8 882.2 10
13 Borated ss +h2o 329.5 351.3 22 41 Manifold
(30%void)
882.2 887.2 4
14 vv-case 351.3 357.3 6 42 void 887.2 889.7 1
15 void 357.3 359.8 1 43 vv-case 889.7 895.7 6
16 Manifold
(30%void)
359.8 364.8 4 44 Borated ss +h2o 895.7 958.7 63
17 Blkt
(ss:88,h2o:10.8)
364.8 375 10 45 vv-case 958.7 964.7 6
18 Blkt
(ss:88,h2o:10.8)
375 385 10 46 t-shield 964.7 970.2 2
19 Blkt 385 390 5 47 void 970.2 1029. 1
(ss:80,h2o:18)
9
20 Blkt 390 394.1 5 48 Inter Coil 1029. 1041. 12
(ss:80,h2o:18)
Structure 9 9
21 blktwall 394.1 395.9 2 49 void 1041. 1349. 5
9 7
22 void 395.9 396.8 1 50 cryostat 1349. 1354. 5
7 7
23 f/w 396.8 399.4 3 51 void 1354. 1366. 1
(ss:65,h2o:31.5)
7 7
24 f/w-ss2 399.4 401.7 2 52 bioshield 1366. 1566. 85
7 7
25 f/w-cu 401.7 403.9 3 53 Void 1566. 1600. 1
7 0
26 be 403.9 404.9 1 54
27 scrape-off 404.9 419 1 55
28 plasma 419 619 4 56
Total Mesh Number 400
Nuclear Analysis Report Page 18

ITER G 73 DDD 2 01-06-06 W0.1
Note: In the case of TF coil nuclear response calculation, Inter Coil Structure was replaced with TF coil and
cryostat and bioshield were eliminated, changing also numbers of zones (53 to 56) and spatial mesh points (400
to 390).
3.2 Radiation Fluxes and Nuclear Heat Distribution during
Operation
3.2.1 Radiation fluxes
Figures 3.2-1a and 3.2-1b show neutron and gamma ray fluxes in the inboard and the
outboard regions during nominal operation. The average 14 MeV neutron current at the first
wall (neutron wall load) is 0.55 MW/m 2 and resulting inboard average is 0.431 MW/m 2 and
outboard 0.641 MW/m 2 .
The neutron wall load distribution obtained by 3-D calculation 1 shows that the peak values of
0.78 MW/m2 in outboard and of 0.59 in inboard. Then estimated peaking factors are 1.21 =
0.78/0.641 in outboard and 1.37 =0.59/0.431 in inboard, respectively.
1.E+15
1.E+14
1.E+13
1.E+12
1.E+11
1.E+10
1.E+09
1.E+08
1.E+07
1.E+06
1.E+05
1.E+04
1.E+03
1.E+02
TFC
VV
14MEV/N.
0.1MEV/N
TOTAL/N.
TOTAL/G.
1.E+01
200 250 300 350 400 450 500
Distance from the Torus Axis (cm)
Figure 3.2- 1a Neutron and Gamma-ray fluxes during operation
Inboard Side
1 G. Ruvutuso and H. Iida ,NAG-156," Neutron Wall Loading with a Non-inductive Operation Plasma"
Nuclear Analysis Report Page 19
Blkt

ITER G 73 DDD 2 01-06-06 W0.1
1.E+15
1.E+14
1.E+13
1.E+12
1.E+11
1.E+10
1.E+09
1.E+08
1.E+07
1.E+06
1.E+05
1.E+04
1.E+03
1.E+02
Blkt
14MEV/N.
0.1MEV/N
TOTAL/N.
TOTAL/G.
1.E+01
800 850 900 950 1000 1050 1100
Distance from the Torus Axis (cm)
Nuclear Analysis Report Page 20
VV
TFC
Figure 3.2- 1b Neutron and Gamma-ray fluxes during operation
Outboard Side
Table 3.2-1 shows fast neutron fluence in the inboard TF coil leg. Total operation time was
assumed to be consistent to the 0.5 MWa/m2 of fluence at the first wall. The expected
fluence in the TF coil insulator and in the winding pack is smaller than the design limit by
factors of about 2 and 40, respectively.
Table 3.2-1. Fast neutron* fluence in TF Coil inboard Leg
Design Limit 1-D Correction for multidimension**
r-Epoxy Insulator 5.0E+17 7.83E+16 2.58E+17
(n /cm2/s)
First layer of
Winding Pack
(n/cm2/s)
1.0E+19
(Nb3Sn)
*Neutron E > 0.1MeV
**Gap effect: 2, Flexible Joint: 1.2, Peaking in inboard:1.37
3.2.2 Nuclear heat distribution
6.91E+16 2.28E+17
Figure 3.2-2a and 3.2-2b show the nuclear heat distribution in the blanket and vacuum vessel.
It should be noted that the 1-D calculation (especially “annulus model”) gives significant
overestimation of the nuclear heating near the plasma because of the neutron flux grazing
(very small incident angle) components which do not exist in real tokamak geometry. Table

ITER G 73 DDD 2 01-06-06 W0.1
3.2-2 shows the integral nuclear heat in the blanket and the vacuum vessel indicating that the
energy increase by nuclear reaction is about 1.47 = 588/400. Table 3.2-3 shows nuclear heat
estimation in the TF coil inboard leg.
Table 3.2-2 Nuclear Heating in Blanket and Vacuum Vessel
Inboard Outboard Total
3-D
(MW) (MW) (MW) modification
First Wall
(8.1 cm)
88 229 317
Blanket* 69 198 267 585(F/W+Blkt)
Vacuum Vessel 0.66 2.2 2.9 9.1**
*Includes divertor, all plugs in ports
** factor2: Gap effect between inboard blanket modules
factor 3: Gap effect between outboard blanket modules
factor 1.15: Gap effect between blanket and divertor (for the both of inboard and outboard)
Nuclear Heat per
unit Height
(W/cm)
Table 3.2-3. Nuclear Heating in TF Coil inboard Leg
Case r-epoxy W.P.** Total Total
1.74 0.10 1.86 3.70
Total Nuclear
Heating in the
Leg* (kW)
*effective height: 610 cm
1.06 0.061 1.13 2.25
**W.P. stands for Winding Pack
When the effects of the blanket module support “flexible joint” (1.2), gaps between the
blanket modules (2.0), gap between the divertor cassette and the blanket modules (1.15) and
vacuum vessel heterogeneity effect (1.2) are taken into account, the estimate of total nuclear
heating in the TFC inboard leg become as follows.
2.25 (kW) x 1.2 x 2.0 x 1.15 x 1.2 = 7.5 kW
Table 3.2-4 gives the absorbed energy in the inboard TF coil insulator assuming 0.5
MWa/m2 of average fluence at the first wall. The absorbed energy in the insulator will be
smaller than the design limit by a factor of ~3.
Nuclear Analysis Report Page 21

ITER G 73 DDD 2 01-06-06 W0.1
Absorbed
Energy (MGy)
Table 3.2-4 Absorbed energy in the inboard TF coil insulator
Design Limit One-D 3-D*
modification
10 0.67 3.2
*Gap effect: 2, Flexible Joint 1.2, Poloidal peaking in inboard 1.37, Local peaking by partial 3-D 1.2, V.V.
heterogeneity 1.2.
1.E+01
1.E+00
1.E-01
1.E-02
1.E-03
1.E-04
1.E-05
Be -T
1.E-06
250 300 350 400 450
Distance from the Torus Axis (cm)
F/W-Cu
BLKT
Nuclear Analysis Report Page 22
VV
TFC
R-EPOXY
Figure 3.2-2a Inboard nuclear heat distribution

ITER G 73 DDD 2 01-06-06 W0.1
1.E+01
1.E+00
1.E-01
1.E-02
1.E-03
1.E-04
1.E-05
1.E-06
1.E-07
1.E-08
800 850 900 950 1000 1050 1100
Distance from the Torus Axis (cm)
Be -T
F/W-Cu
BLKT
Nuclear Analysis Report Page 23
VV
TFC
R-EPOXY
Figure 3.2-2b Outboard nuclear heat distribution
Table 3.2-5 shows nuclear heating (or absorbed dose) rate in the thermal shield behind the
vacuum vessel. This analysis is important since radiation could affect the emissivity of
materials. Dose on the outboard thermal shield is much smaller than that on the inboard in
this 1-D analysis result. However at some part, for example near the NBI port, it can be
higher than those shown in this table and will be assessed by 3-D analysis in future.
Table 3.2-5 Nuclear heating (or absorbed dose) rate in the thermal shield behind the
vacuum vessel.
Nuclear Absorbed Absorbed Absorbed
heating Dose dose dose(Gy)
(w/cc) Rate (Gy/s) (Gy) *3Dmodification
Inboard Neutron 1.51E-5 1.9E-3 5.4E+4 2.6E+5
Gamma-ray 1.41E-4 1.8E-2 5.0E+5 2.4E+6
Total 1.57E-4 2.0E-2 5.5E+5 2.6E+6
Outboard Neutron 8.09E-8 1.0E-5 2.8E+2 -
Gamma-ray 6.02E-7 7.7E-5 2.1E+3 -
Total 6.83E-7 8.8E-5 2.4E+3 -
*Gap effect: 2, Flexible Joint 1.2, Poloidal peaking in inboard 1.37, Local peaking by partial 3-D 1.2, V.V.
heterogeneity 1.2.

ITER G 73 DDD 2 01-06-06 W0.1
3.3 Helium production rate
Figure 3.3-1a and 3.3-1b show the helium production rate distribution in inboard and
outboard. At the critical point of the vacuum vessel surface, helium production rates are 0.10
appm for the inboard and 0.13 for the outboard when average fluence at the first wall is 0.3
MWa/m 2 . From 3-D calculation (JP HT) 1 it was estimated that the peaking factor caused by
the 2 cm of gaps between blanket modules is about six. As a consequence the expected
maximum helium production rates are 0.6 appm for the inboard and 0.8 appm for the
outboard.
Those values are smaller than the design limit values ( ~ 1 appm for thick welding and ~3
appm for thin welding).
Helium Production in 316SSig (appm)
1.E+02
1.E+01
1.E+00
1.E-01
VV
Blkt
1.E-02
340 360 380 400 420
Distance from the Torus Axis (cm)
Figure 3.3-1a Helium production rate distribution in the inboard (in SS)
Nuclear Analysis Report Page 24
F/W
1 S. Satoh, J. Ohmori "Report on Nuclear Response Analysis in TFC and VV (V.1.2)

ITER G 73 DDD 2 01-06-06 W0.1
Helium Production in 316SSig (appm)
1.E+02
1.E+01
1.E+00
1.E-01
F/W
1.E-02
840 860 880 900
Distance from the Torus Axis (cm)
Figure 3.3-1b Helium production rate distribution in the outboard (in SS)
3.4 Damage in the Blanket, Vacuum Vessel and Divertor Materials
The atomic displacement, helium, and hydrogen production levels have the most impact on
the neutron-induced effects in structural alloys. In particular, the damage accumulated in
materials, and preloading relaxation induced by radiation, is an important design constraint
for the development of the separable first wall fasteners and blanket attachments.
Different methods and tools have been used to estimate radiation streaming through access
holes and gaps, and resulting nuclear responses in the first wall, blanket, vacuum vessel and
the divertor. Some of the peculiarities of the damage production are described below.
3.4.1 Spectral Effects
Principally, the damage cross sections 1 depend on the material and the neutron spectrum.
Figure 3.4-1 compares the neutron flux, damage production rate, and Helium-production rate,
integrated over the neutron spectrum in the ITER steel/water blanket as a function of its
upper energy 2 .
1 M. E. Sawan, Calculational Benchmark Results with the New Multi-Group Processed Library. Fusion
Technology Institute, UW. IAEA Advisory Group Meeting on “Extension and Improvement of the FENDL
Library for Fusion Applications”, 3-7 March 1997, Vienna, Austria.
2 V. Khripunov and H. Iida, Damage in Bimetallic Studs and Inconel Bolts. G 16 MD 281 00-05-26 F 1 (NAG-
158-26-05-00), Garching, May 2000.
Nuclear Analysis Report Page 25
Blkt
VV

ITER G 73 DDD 2 01-06-06 W0.1
Relative Values
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15
Energy, MeV
F(

ITER G 73 DDD 2 01-06-06 W0.1
Figure 3.4-2 Inboard Radiation Damage (DPA) Distribution
1.E+01
1.E+00
1.E-01
1.E-02
1.E-03
1.E-04
1.E-05
1.E-06
1.E-07
1.E-08
Flexible Joint bolt
F/W fastener bolt
1.E-09
800 850 900 950 1000 1050 1100
Distance from the Torus Axis (cm)
Nuclear Analysis Report Page 27
Be
Fe
Cu
W
Figure 3.4-3 Outboard Radiation Damage (DPA) Distribution
They show the dpa-values in different materials when they are placed in an arbitrary position
in the blanket or the vacuum vessel and are exposed by the radiation fluxes with energy
spectrum at that point. The averaged neutron fluence at the first wall is assumed to be 0.5
MWa/m 2 , giving local peak of 0.7 MWa/m 2 on the outboard and 0.54 MWa/m 2 on the
inboard, respectively.
More detailed dpa-distributions in the blanket are illustrated in Figure 3.4-4. It compares also
the damage accumulated in the first wall and blanket materials as well as in the blanket
attachment materials. For a convenience these distributions may be described by the
exponential law with the exponent attenuation factor of ~ - 0.12 cm -1 .

ITER G 73 DDD 2 01-06-06 W0.1
10
1
0.1
0.01
0 5 10 15 20 25 30 35 40 45 50 55
Distance from the First Wall, cm
Nuclear Analysis Report Page 28
Ti6AlV
Inconel
Figure 3.4-4 One-D Radial Distributions of Damage
in the Blanket and Attachment Materials (Neutron Fluence 0.5 MWa/m 2 )
Atomic displacements in titanium alloys resulting from fusion neutron irradiation is ~10-30
% higher than that in steel. The damage in other materials are nearly the same for the same
locations in the blanket depth as that in steel: in Inconel by ~5-15 % higher, in Cu-alloy is
slightly (~ 0-10 %) higher.
3.4.3 Peaking Factors
The peaking factor for damage at the bolt end surface does not exceed ~1.3, in comparison
with the value in the bulk shield at the same distance (~33 cm) from the first wall 1 2 . The
peaking factor, estimated for the 13 mm hole to access the bimetallic stud at the distance ~70
mm from the first wall inner surface, is about ~1.1 (See reference 3 ).
3.4.4 Damage in the First Wall Fastener Materials
The first wall fastener assemblies in blanket with mechanically attached first wall (See Figure
3.4-5) are located at a distance of ~ 6-15 cm from the inner surface of the plasma chamber
and attach the separate first wall panels to the shielding blanket blocks.
1 R. T. Santoro, H. Iida and V. Khripunov, MCNP Calculations of Nuclear Responses in the ITER Shielding
Blanket Flexible Attachment. G 73 RI 19 97-11-17 F1 (NAG-15-12-17-96), Garching, December 1996.
2 R. Plenteda, H. Iida, D. Valenza and R. T. Santoro, 3-D Monte Carlo Analysis of the Peaking Effect An
Improved Solution for the Equatorial Inboard Flexible Attachment. NAG-115-18-10-98–rev-1, Garching,
October 1998.
3 S. Sato et al., Blanket and Flexible Joint Analysis. Report at the Progress Meeting on Neutronics Design
Tasks and the R& D Task T426. Garching JWS, 4-5 July, 2000.
Cu
SS
Be

ITER G 73 DDD 2 01-06-06 W0.1
SHIELDING
BLOCK
~ 2.3
dpa
~ 3.4
dpa
~ 1.5
dpa
~ 1.1
dpa
~ 2.9
dpa
~ 5.3
dpa
BIMETALLIC STUD
(INCONEL+COPPER)
THREADED
BUSH
Figure 3.4-5 Damage Expected in Some Elements of the First Wall Fastener
(Neutron Fluence 0.5 MWa/m 2 )
The axial symmetry of these systems and access hole geometry require the use of twodimensional
models. Based on the dimensions and material compositions from reference 1 ,
two (R, Z)- models of the attachment assemblies were developed for discrete ordinates
radiation transport calculations. In the axial (Z) direction they include the plasma region, the
multilayer first wall, the 80%SS/20%H 2 O - shielding blanket, and the vacuum vessel. In the
radial (R) direction all elements of the first wall fastener or the blanket attachment assemblies
with their access holes, the steel blanket module walls were modelled as cylinders. The
simplified blanket structure was used on the distance more than 50 cm from the symmetry
axis to achieve the appropriate boundary conditions.
The damage accumulated in different elements of the first wall fasteners (such as the
threaded bush and bimetallic stud made of Inconel with a copper rod) calculated in a 2D
geometry 2 are given in Table 3.4-1 as a function of distance from the inner surface of the
first wall. These data are positioned also in Figure 3.4-5 relating to the first wall.
The maximum damage in the Inconel threaded bush is ~3.4-2.9 dpa on its axis. A factor of
two lower values are expected in the lower part of the bolt shank. It should be noted that the
ratio of the He-gas production (appm) to the dpa-production rate at the first wall is about 840
for the Beryllium coverage that is much higher than that is typical for steel (~13).
1 Design Drawings No. 16_0422.eps (FW-Fastener), 16_0416.eps (Bolt for Blanket Attachment), modified by
Th. Vollman, Garching, 19 May, 2000.
Nuclear Analysis Report Page 29
~ 1.6
dpa
~ 1.0
dpa
~ 1.4
dpa
~ 2.7
dpa

ITER G 73 DDD 2 01-06-06 W0.1
Table 3.4-1 Damage in the First Wall Fastener Surrounding
Distance from inner Material Location Damage, dpa
FW surface, cm
per 0.5 MWa/m 2
0 Be Be-coverage of the FW 1.7
1 Be Be-coverage of the FW 1.4
Cu Cu-layer of the FW 5.5
~ 2 Cu Cu-layer of the FW 4.7
~ 5.2 Inconel Threaded Bush 3.4
6.6 Inconel Bimetallic Stud 2.9
Cu Copper rod 2.9
Steel First Wall Panel 2.7 (2.6) *)
8 Inconel Threaded Bush 2.4
Steel Shielding Block 2.3 (2.2) *)
12 Inconel Bimetallic Stud 1.5
Cu Copper rod 1.4
Steel Shielding Block 1.4 (1.3) *)
*) Values in brackets are given for the points on the distance ~ 10-15 cm or more
from the fastener symmetry axis.
3.4.5 Vacuum Vessel
The outboard blanket provides two order of magnitude attenuation for the fast neutron flux.
Therefore, the damage production level in the vacuum vessel attenuates exponentially
beginning from ~0.02 dpa in the front steel layer. This “dose” is low and will not result in
significant property changes of the structural materials.
3.4.6 Damage Function of Neutron Fluence
All the above values are the peak neutron-induced damage to the first wall, blanket and
vacuum vessel. They are normalised to the maximum local neutron fluence of 0.5 MWa/m 2
(~ 3.5 10 21 neutrons (En > 0.1 MeV) per cm 2 ). This is the local maximum expected in the
outboard first wall at the end of the DT-operation campaign with the nominal average
neutron fluence of 0.3 MWa/m 2 , that corresponds to the total burn time ~ 2 10 7 s, or about
0.63 full power year (FPY).
The dpa-values and the spatial distributions presented may be used for scaling to other local
fluence and bolt positions.
3.4.7 Displacements and other responses in the Divertor
For a comparison, displacement levels as well as helium production and power density in the
divertor plasma facing and structural materials 1 are given in Table 3.4-2.
1 V. Khripunov, Irradiation Conditions for the ITER-FEAT In-Vessel Materials. Report at the Technical
Meeting “ Materials for In-Vessel Components”, Garching, 31 January - 4 February 2000.
Nuclear Analysis Report Page 30

ITER G 73 DDD 2 01-06-06 W0.1
Table 3.4-2 Peak Nuclear Responses in the Divertor Cassette Components
(Total Burn Time ~ 2 10 7 s, or 0.63 FPY)
Neutron Wall
Loading,
MW/m 2
Effective
Fast Neutron
Fluence, cm -2
Materials W/cm 3
dpa He appm
Outer Vertical Target
0.42 1.8 10 21
W 12 ~ 0.7 0.4
Cu 2.3 ~ 1.7 16
SS 1.4 0.5 6
~ 0.17 8 10 20
CFC 2.3 ~ 0.7 ~ 230
Attachment and Cooling Pipes Connections
SS 0.5 0.2 1.3
Dome Surface Layers
0.43 1.8 10 21
W 12 0.6 ~ 0.35
Cu ~ 4 1.7 15
SS ~ 2 ~0.35 ~ 3.5
The neutron wall loading peaks ~0.4 MW/m 2 in the central dome and the top of the outer
vertical targets which have the largest view of the plasma. That is ~ 2 times lower than the
first wall maximum. Therefore, the nuclear parameters in the divertor region, in general, are
lower than in the first wall.
The tungsten plasma facing components at the top of the vertical targets and the dome surface
experience relatively high level of the heating and damage.
They are moderate in the cassette body. Heating and damage at the coolant connection and
attachment locations drop to about 0.5 W/cm 3 , 0.2 dpa and ~ 1 He appm per 0.63 FPY,
receptively. As a result, pipe rewelding should be feasible.
3.5 Dose Rates during Operation and Shutdown
Figure 3.5-1 shows the operation dose rate distribution obtained by a 1-D calculation. During
machine operation, the dose rate inside the bio-shield is too high for personnel access. Dose
rate behind the bio-shield is low enough (~1 μSv/h ) for personnel access in this figure.
However, practically, it may be impossible to access there because of radiation streaming
through the many penetrations in the bio-shield. (Such access during operation will also be
ruled out by the presence of varying magnetic fields.) Examples of typical cases of such
analyses are presented in the chapter 7.
Nuclear Analysis Report Page 31

ITER G 73 DDD 2 01-06-06 W0.1
1.E+15
1.E+14
1.E+13
1.E+12
1.E+11
1.E+10
1.E+09
1.E+08
1.E+07
1.E+06
1.E+05
1.E+04
1.E+03
1.E+02
1.E+01
1.E+00
1.E-01
Plasma
Blanket
&
Vacuum
Vessel
design limit for the space inside cryostat
acc
design limit for the space behind bioshield
600 800 1000 1200 1400 1600
Distance from the Torus Axis (cm)
Bioshield
Figure 3.5-1 Dose Rate Distribution During Operation
In the space inside the cryostat, limited personnel access is expected for machine
maintenance. At the locations where personnel should have access, the dose rate should be
less than 100 μSv/h about 2 weeks (~ 10 6 s) after shutdown.
The operation scenario, “M-DRG1” and is supposed to be used for dose rate calculation after
shutdown (see section 2.3). The accumulated fluence of each year was simulated by
lowering the fusion power assuming continuous operation through the year. Only 6days high
load factor campaigns at the end of the former and latter 10 years were simulated by 36 one
hour pulses with 3 hour dwell time among them. The integrated fluence is 0.3 MWa/m2 at
the first wall (averaged over the whole first wall surface).
The results of 1-D calculation is shown in Figure 3.5-1. Generally the dose rate 10 6 sec after
shutdown is lower than that during operation by 5 – 6 orders of magnitude and is 10 – 100
micro Sv/h outside the vacuum vessel ignoring radiation streaming through the penetrations.
For actual dose rate estimation inside the cryostat requires 3-D analyses fully taking account
of streaming through various penetrations. The detail of 3-D analyses are reported in the
chapter 5 and 6.
The calculated dose rate outside the bioshield is very low and is even below the natural
background (~ 0.1 micro Sv/h) as far as there is no penetrations in the bioshield. If we can
provide one order of magnitude attenuation in spite of existence of penetrations, the dose rate
outside bioshield at 10 6 sec after shutdown can be lower than the design limit of 10 micro
Sv/h, supposing that the design limit of 100 micro Sv/h inside the cryostat is satisfied.
When personnel have to access outside the bioshield sooner than that, a little bit larger
attenuation is required as shown in Figure 3.5-1 and 3.5-2. This is because of Na-24
Nuclear Analysis Report Page 32

ITER G 73 DDD 2 01-06-06 W0.1
(T1/2=15.02h) which is produced in the concrete of the bioshield. By adding small amount of
boron in the bioshield, this effect is significantly reduced (see Figure 3.5-2).
Dose Rate (microSv/h)
1.E+11
1.E+10
1.E+09
1.E+08
1.E+07
1.E+06
1.E+05
1.E+04
1.E+03
1.E+02
1.E+01
1.E+00
1.E-01
1.E-02
1.E-03
600 800 1000 1200 1400 1600
Distance from the Tirus Axis (cm)
Nuclear Analysis Report Page 33
0 s
1.0e+4 S
1.0e+5 S
1.0e+6 S
Figure 3.5-1 Dose Rate Distribution after M-DRG1 Operation
Dose Rate (microSv/h)
1.E+04
1.E+03
1.E+02
1.E+01
Inter-Coil
Structure
Required
attenuation is
1.E+00a
little larger
than one
order of
1.E-01magnitude
Blanket
VV
Bio-shield
1.E-02
1300 1350 1400 1450 1500 1550
Distance from the Tirus Axis (cm)
0 s
1.0e+5 S
1.0e+6 S
0S+boron
1.e+5S+boron
1.e+6S+boron
Required
attenuation is
two order of
magnitude
Line of oneorder
of
magnitude
attenuation
Figure 3.5-2 Detailed Dose Rate Distribution around the Bioshield

ITER G 73 DDD 2 01-06-06 W0.1
3.6 Decay Heat
The operation scenario for assessing decay heat is more conservative than the reference
scenario for shutdown dose estimate, since it will be used for the accident analyses. A “ New
SA1” scenario was defined by the safety group for decay heat analyses 1 and used only for
decay heat analysis in this report. The SA1 lasts for 10 years and one month and
approximated in actual calculations as shown in Figure 3.6-1. The integrated fluence is 0.5
MWa/m2 in this definition differing from that in dose rate analyses, which prefer more
practical (less conservative) assumption. Figure 3.6-2 show the decay heats for the SA1
scenario. The decay heat in short period after shutdown is dominated by the last one hourlong
pulse, one day after shutdown by the last 3 day pulsing and after that by rather high load
factor operation in the latter 5 year operation.
New-SA1 with 0.5 MWa/m2
100%: 0.57 MW/m2
0.0%
5 years
Figure 3.6-1 New SA1 operation scenario for decay heat calculation (ITER)
1.E+02
1.E+01
1.E+00
1.E-01
1.E-02
1.E-03
4.540%
5 years
2.528%
5 years
1.E-04
1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 1.E+08 1.E+09
1 GENERIC SITE SAFETY REPORT (GSSR)
9.697%
5 years
Time after Shutdown (sec)
30 %
27
days
5y-1
5y-2
27day
21 pulse
3600 sec
total
3days pulse operation
1 hour on power
2.33 hours dwell time
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 3.6-2 ITER decay heat (new SA1 with 0.5 MWa/m2)
Note: This calculation employed copper rich First Wall option which gives 30 % higher values during the first
one hour.
1.E+02
1.E+01
1.E+00
1.E-01
ITER-SA1(0.5MWa)
ITER-SA1(0.3MWa)
98FDR-M5a(0.3MWa)
1.E-02
1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 1.E+08 1.E+09
Time after Shutdown (sec)
Figure 3.6-3 Comparison of decay heat in reference case with other optional scenario
Figure 3.6- 3 shows the comparison of ITER decay heat (SA1(0.5MWa)) with those of
shorter operation scenario (0.3 MWa) and of the previous design (‘98FDR-ITER). The
reduced fluence case (SA1(0.3MWa)) gives similar value as the reference one up to one day
after shutdown, but less (~ 60%) values after that. In comparison with the previous design,
labelled “M5a.”, the decay heat for ITER gives much lower values in short period after
shutdown because of mainly its lower (1/3) fusion power.
3.7 Activation of air outside the Cryostat and Bioshield 1
Possible concerns associated with Ar41(T1/2 = 1.83 h) production in air in the Tokamak
building are the followings
• It might cause excessive irradiation of personnel in the building.(mpc ; ~0.1 Bq/cc)
• Ar-41 concentration in the exhaust air from the building could be higher than the
allowable concentration of Ar-41 giving higher concentration at the site boundary of
ITER than the permissible concentration to the public.(0.0005Bq/cc at sit boundary).
1 H. Iida ,NAG-160”Activation of Air inside the Building producing Ar-41”
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ITER G 73 DDD 2 01-06-06 W0.1
The HVAC system of the building should be designed so that the above problems are
avoided. In this section the preliminary estimation of Ar-41 concentration in the static air in
the building are provided.
3.7.1 Location of the Assessment and calculated production rate
Estimation has been made with an one-dimension calculation with annulus model shown in
the Figure 3.7-1 which is basically same as Figure 3.1-1 in the section 3.1. The concentration
of Ar-41 is calculated in two locations (A and B in Figure. 3.7-1) which are inside and
outside the bio-shield. In the actual plant, there will be some location which is under different
geometrical condition from 1-D analysis, for example NBI pit, but should not give a much
larger value of Ar-41 production rate than the above location A (inside bio-shield).
A B
CS
TF
VV
Blkt Blkt
Plasma
Figure 3.7-1 Annulus model Geometry
Ar-41 is produced through Ar-40 (n,γ) Ar-41 reaction. The Ar –40 concentration in air was
assumed to be 0.93 %. 1
Table 3.7-1 shows the calculation result with assumption of 10 hours continuous operation.
The production rate is naturally same as the saturated decay rate shown in this table. The Ar-
41 production outside the bioshield is so small ( ~6 orders of magnitudes) that it can be
neglected in the following estimation of Ar-41 concentration in the Gallery space.
Table 3.7-1 Ar-41 Production in the air (Bq/cc)
Location A (inside the Bio-shield) Location B (outside the Bio-shield)
2.44E+01 1.27E-05
3.7.2 Ar-41 concentration in the air
Using the Ar-41 production rate value in Table 3.7-1 and assuming the followings 2 , the Ar-41
concentration in the air has been estimated as shown in Figure 3.7-2.
1 taken from the home page of Geosciences at Emory University in Atlanta
2 private communication from the tritium system group in Naka JCT
Nuclear Analysis Report Page 36
VV
Inter Coil
Structure
Cryostat
Bio-Shield

ITER G 73 DDD 2 01-06-06 W0.1
• Operation scenario: infinite pulsing with 400 sec of on power and 1200 sec of dwell time.
• The volume of the space between the cryostat and the bioshield: 5300 m 3
• The volume of the gallery area: 65000 m 3
• Ventilation speed of the space between the cryostat and the bioshield: 100 %/day
• Ventilation speed of the gallery space: 100 % /day
1.E+01
1.E+00
1.E-01
1.E-02
1.E-03
Cryo.-Bio.-400s
Garally-400s
0.0E+00 5.0E+03 1.0E+04 1.5E+04 2.0E+04 2.5E+04 3.0E+04 3.5E+04 4.0E+04
Time (s)
Figure 3.7-2 Ar-41 Concentration in the space between the crystat and the bioshield and
in the Gallery
The concentration in the space between the cryostat and the bioshield is about 6 Bq/cc which
is 1/4 of the production rate. The concetration in the gallery space is about two orders of
magnitude smaller than that inside the bioshield because of the following two factors :
• Volume ratio of the two space : 5300 / 65000 = 0.082
• The ratio of the ventilation speed and Ar-41 decay rate: (0.693/24) / (0.693/1.83)= 0.076
As a conclusion, the estimated Ar-41 concentration (~ 0.05 Bq/cc) will be lower than the
MPC for workers (0.1 Bq/cc). Only 2 orders of magnitude reduction of Ar-41 concentration
from gallery space to the site boundary is required for satisfying the limit for site boundary
(0.0005Bq/cc).
It should be noted that the Ar-41 production outside the bioshield plug may be taken into
account if the plug will provide radiation attenuation by less than two order of magnitude.
Nuclear Analysis Report Page 37

ITER G 73 DDD 2 01-06-06 W0.1
3.8 Water Coolant Irradiation
As reported in references 1 2 3 , the coolant water in the ITER will be activated by high energy
neutrons via the 16 O(n, p) 16 N and 17 O(n, p) 17 N reactions, as it flows through the cooling
channels located in the plasma facing components of the blanket and the divertor.
Two main problems associated with 16 N and 17 N radionuclides, carried by irradiated water
flowing at the velocities of 2-6 m/s out of the reactor core, were identified during the ITER
EDA:
(1) the “hot” outlet pipes are a strong source of high energy 16 N-decay photons (~6 or
7 MeV) inside the cryostat, resulting in radiation problems in cryogenic components
during operation, and behind the biological shield in the primary coolant loop
environment ~ 1 minute after irradiation;
(2) the internal 17 N-decay 0.9-MeV neutron source in the water affects the cooling
pipe residual activity.
The intensities of the short lived 16 N gamma-ray source (T1/2= 7.12 s) and 17 N fast neutron
source (T 1/2= 4.12 s) in irradiated flowing water depend strongly on the details of the water
flow path, and the fast (>10-MeV) neutron flux distribution in the plasma region. The
nitrogen content is changed by exposure to the neutron flux and is a function of irradiation
time and delay time after irradiation before exiting the blanket and the vacuum vessel. That
is why the specific water radioactivity was estimated by accounting water velocity, coolant
pipe dimensions, residence and delay times in the first wall, interior blanket and divertor
structures, manifolds and outlet coolant pipes. Neutron wall loading and mass flow rate
distributions have been also included in the analysis.
3.8.1 Critical Locations
There are locations in the reactor ex-vessel space where the 16 N decay gamma-ray heating is
critical (See Figure 3.8-1). These include:
- the main outlet water flow after cooling inboard and outboard blanket modules in the
upper ports, where the “hot” return pipes are surrounded by 20-cm-thick port walls.
Outside the ports (but inside the cryostat) they bend toroidally, vertically and finaly
radialy and are screened only by a thin 9-mm steel transition thermal shield. So the
main 16 N decay gamma-ray source inside the cryostat is concentrated in these region.
- the outlet water pipes of the diagnostic or limiter cooling systems exiting through the
upper or mid-plane ports where they are contained in the port walls; and
1 V. Khripunov and R.T. Santoro, Radionuclide Production in the ITER Coolant Water, ITER Report No.G 30
RI 1 96-07-10 W1.1. (IDoMS No. NA/NAG-50). Garching JWS, July 1996.
2 R. T. Santoro, V. Khripunov, H. Iida, R. R. Parker, Radionuclide Production in the ITER Water Coolant. 17th
Symposium on Fusion Engineering, San-Diego, California, 6-10 October, 1997.
3 V. Khripunov, R. T. Santoro, H. Iida, R. Parker, G. Janeschitz, R. Plenteda, Activation of Water Coolant in
ITER. Proceedings of the 20th Symposium on Fusion Technology, Marseille, France, 7-11 September, 1998.
Vol.2, pp 1405-1408.
Nuclear Analysis Report Page 38

ITER G 73 DDD 2 01-06-06 W0.1
- the divertor coolant outlet pipes that pass through the divertor pumping and
maintenance ports and nearby the cryostat are not shielded.
Figure 3.8-1 Cooling Circuits to the Blanket, Diagnostic Plugs and Divertor
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ITER G 73 DDD 2 01-06-06 W0.1
3.8.2 Nitrogen in the Blanket Water Coolant
3.8.2.1 Residence Time
Calculations of the water residence and delay time in the shielding blanket and manifolds
were performed based on the data given in references 1 and 2 , using detailed water flow
parameters for the blanket cooling system. They include, in particular, a collector effect, i.e.
water velocity changes along a manifold length, introducing an additional delay, depending
on the flow path.
There are different blanket module types in ITER. They are arranged in several groups that
are connected by different sets of manifolds. Eight outlet coolant pipes from manifolds pass
through each upper port (Figure 3.8-2).
Figure 3.8-2 Blanket Coolant Pipes in the Upper Port
All cooling pipes have the same inner diameter ~ 65 mm, and the wall thickness ~5.3 mm.
The water velocity is ~ 5-6 m/s and the mass density ~ 0.94 g/cm 3 . The length of the pipes
inside the ports varies from ~ 4 to 5 m. The rest of the pipes (~ 11.3 m) is located outside the
port plugs. Nine bundles of feed and return pipes penetrate the cryostat and the 2-m-thick
biological shield to the main collector and the primary heat exchanger in the upper pipe
chase. (See Figure 3.8-1).
The calculated water irradiation time and delay time in the coolant flow path between each
module and the return pipes exit the ports are given in Table 3.8-1.
1 Plant Description Document, Section “3.3 Cooling Water”. G A0 FDR1 00-11-16 W 0.1. Garching,
December 2001.
2 S. Keijers, W. Vanhove, BELGATOM. Private communication, 27 October, 2000. See also: “Development of
an ATHENA Model for the ITER-FEAT FW/Blanket Cooling system.” Framework Contract –Task Order
NET/93-851 FU, BELGATOM, August 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
The total residence time of the cooling water in the blanket is about 9 s with the water
residing in the high neutron flux regions (at the first wall) for about 2 s.
Module
No.
Table 3.8-1 Water Residence Time in the Blanket Modules,
Manifolds and Return Pipes in Upper Ports
Number Flow rate Residence time, s
of
modules
per loop
per one
module,
kg/s
Module inlet
- first wall
First wall Blanket body
- module
outlet
Manifolds
- return pipes
in upper ports
1 6 3.1 1.7 5.4 16.0 12.0
2 6 6.5 0.8 2.5 9.4 6.5
3 6 8.5 0.6 2.0 7.2 4.7
4 6 9.0 0.6 1.9 6.8 3.8
5 6 7.7 0.7 2.3 7.9 3.2
6 6 5.2 1.0 3.3 11.8 2.7
7 6 6.7 0.8 2.8 9.5 2.3
8 9 6.1 0.8 2.5 9.2 1.9
9 9 6.1 0.8 2.5 8.8 1.6
10 9 7.1 0.7 2.0 8.1 2.2
11 12 7.6 0.6 2.0 6.9 1.3
12 12 8.2 0.6 1.7 6.6 2.2
13 6 7.8 0.6 1.7 6.3 2.5
14 6 9.6 0.5 1.7 5.7 2.9
15 12 9.0 0.6 1.8 6.5 3.7
16 12 8.7 0.6 2.1 6.7 4.5
17 12 6.7 0.7 2.4 8.3 5.7
average 0.6 1.9 6.9 2.9
3.8.2.2
16 N and 17 N in the Blanket Water Coolant
The radionuclide production rates estimated for the reactor conditions as a function of
exposure times are given in Table 3.8-2.
Table 3.8-2 Nitrogen Production Rates and Nuclear Densities
in the Outlet “Hot” Water Pipes
Module inlet
- first wall
First wall Blanket body
- module outlet
Manifolds
- return pipes
in upper ports
Total
Average residence time, s
0.6 1.9 6.9 2.9 12.3
Nuclear density in the return pipes in upper ports, 1/cm 3 (or Ci/cm 3 )
N-16 1.0 10 8
1.1 10 10
5.0 10 9
- 1.6 10 10 (0.042)
N-17 3.1 10 4
3.6 10 6
2.3 10 6
- 5.9 10 6 (2.6 10 -5 )
Nuclear Analysis Report Page 41

ITER G 73 DDD 2 01-06-06 W0.1
The 16 N production rate was calculated separately for each blanket module. Contributions in
the different modules to the total production rate differ from the average value by 30-50%.
The maximum 16 N nuclear densities arise in the central inboard modules and in the outboard
modules where the neutron wall loading is high. Approximately ~ 90 % of the total 16 N
gamma-ray radioactivity is produced in the water during the cooling of the first wall.
The 16 N in the flowing water decays by nearly 50% while it is inside the modules, manifolds
and outlet pipes. Taking into account the water flow rate distribution, the following
parameters were estimated for 16 N in the blanket coolant outlet pipes at their exit the upper
ports:
16 N-nuclear density ~1.6 10 10 cm -3 ;
specific activity ~0.04 Ci/cm 3 .
The average values for the rest pipes inside the cryostat are lower, ~1.4x10 10 cm -3 , and 0.037
Ci/cm 3 , respectively.
The 17 N production rate and associated nuclear responses are three orders of magnitude lower
than those of 16 N (Table 3.8-2). Due to the shorter half-life, ~50 % of the 17 N decays before
it leaves the blanket and manifolds. The 17 N nuclear density in the outlet manifolds at the
exit from the upper ports is ~1 x 10 7 cm -3 (4.5 x 10 -5 Ci/cm 3 ). The total 17 N neutron source
in the outlet coolant pipes inside the cryostat is negligible in comparison with the main
neutron background.
3.8.3 Coolant Activation in Upper and Mid-Plane Diagnostic Ports
Schematic of separate water cooling circuits to diagnostic components is shown in Figure
3.8-1. The water filled “hot” 65-mm pipes activated during the diagnostic plug cooling passes
directly through the port doors, cryostat and the 2-m-thick biological shield into the
diagnostic and maintenance cells. The pipe length is ~10 m with ~3-5 m located in the midplane
and upper ports. The rest of the pipes (~3-5 m) outside the port plugs are unshielded.
Typical values for 16 N and 17 N at the exit of the port doors and the biological shield in return
pipes after diagnostic component cooling were estimated based on the data in Table 3.8-2 and
the water flow characteristics as it is given in Table 3.8-.3.
They do not exceed essentially the average values estimated for the blanket coolant water
(See Table Table 3.8-2).
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ITER G 73 DDD 2 01-06-06 W0.1
Table 3.8-3 Coolant Parameters and Estimated Nitrogen Densities
in the Diagnostic Plugs
Upper Port
Mid-Plane Port
(Module 9 region) (Module 13/14-region)
Velocity, m/s
Pipe Length, m:
2.0 5.6
- in the port 5 3.2
- in port and Bio-shield
Delay time, s:
10 10
- in the port 2.5 0.6
- in port and Bio-shield 5 1.8
N-16: 1/cm 3
Ci/cm 3
1/cm 3
Ci/cm 3
- in the port 1.1 10 10
0.028 2.1 10 10
0.055
- in port and Bio-shield 8.4 10 9
0.022 1.85 10 10
0.049
N-17: 1/cm 3
Ci/cm 3
1/cm 3
Ci/cm 3
- in the port 2.5 10 6
1.1 10 -5
6.0 10 6
2.7 10 -5
- in port and Bio-shield 1.65 10 6
7.4 10 -6
4.9 10 6
2.2 10 -5
3.8.4 Divertor Coolant
The water resident time in the divertor was estimated in reference 1 . The complicated coolant
flow path in the divertor (Figure 3.8-3) was divided into several regions each having different
fast neutron flux levels, water velocities, path lengths, and resident times. The coolant
channel parameters and calculated resident times are given in Table 3.8-4. The average
residence time in the divertor cassette (from inlet to outlet) is about 44 s.
The nitrogen production rate and nuclear densities were estimated for activated water in
different parts of the divertor, weighting the resident time by the fast neutron flux distribution
and taking into account the radionuclide decay (Table 3.8-5).
Coolant connectors
for the inner and outer
vertical targets
Open dome design
with radiative liner
Inner liner
dome
Inner VT
Outer liner
dome
Outer VT
Nuclear Analysis Report Page 43
Inlet
Outlet
Coolant routing
within the cassette body
Figure 3.8-3 Coolant Connector and Routing in the Divertor Cassette Elements 1
1 Ch. Ibbott, Pivate communication, November, 2000. See also: “2.4.8. Divertor Cooling”, in : “Plant
Description Document”, G A0 FDR 1 00-11-16 W 0.1, Garching, December, 2000.

ITER G 73 DDD 2 01-06-06 W0.1
Table 3.8-4 Water Resident Time in the Divertor Cassette
Length, Total Velocity, Time in, Delay,
m length, m m/s s s
Outer support toroid. 1.4 10.3 0.9 12 32
Vertical support poloid. 0.8 1.0 0.8 31.2
Target Swirl CFC 0.8 8.4 0.09 31.2
upper part 1.4 6.3 0.2 31
Cassette feed cooling
Body channel ~5.2 0.92 6 25.3
Inner support toroid. 0.94 7.14 0.73 9.8 15.5
Vertical support poloid. 0.8 1.0 0.8 14.7
Target Swirl CFC 0.8 8.4 0.09 14.6
upper part 0.94 6.3 0.15 14.5
Cassette feed cooling
Body channel 5.15 0.46 5.5 9
Dome copper (upper) 2.40 6.3 0.4 8.6
support toroid. 1.20 3.0 1.1 2.7 5.9
Cassette
Body
support poloid. 2.41 6.3 0.4 5.5
feed cooling
channel 5.15 0.5 5.5 0
grand total ~ 44
Table 3.8-5 Nitrogen Nuclear Density in the Outlet Pipes
as a Result of Water Irradiation in the Divertor
Nuclear density, cm -3
%
N-16 N-17 N-16 N-17
Outer support toroid. 3.0 10 9
8.3 10 4
4.4 0.6
Vertical support poloid. 1.7 10 8
7.0 10 3
0.2 0.05
Target Swirl CFC 3.9 10 7
1.8 10 3
0.1 0.01
upper part 2.0 10 8
9.1 10 3
0.3 0.06
Cassette feed cooling
Body channel 2.0 10 9
1.1 10 5
2.9 0.76
Inner support toroid. 1.2 10 10
1.2 10 6
18 8.3
Vertical support poloid. 4.7 10 8
6.2 10 4
0.7 0.43
Target Swirl CFC 1.1 10 8
1.6 10 4
0.2 0.1
upper part 6.7 10 8
9.4 10 4
1 0.65
Cassette feed cooling
Body channel 9.6 10 9
1.6 10 6
14 11
Dome copper (upper) 3.6 10 9
7.6 10 5
5.2 5.2
support toroid. 1.3 10 10
2.9 10 6
18 20
support poloid. 1.2 10 9
3.1 10 5
1.8 2.2
Cassette feed cooling
Body channel 2.3 10 10
7.3 10 6
33 50
Total 6.9 10 10
1.5 10 7
100 100
(Ci/cm 3 ) (0.18) (6.5 10 -5 )
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ITER G 73 DDD 2 01-06-06 W0.1
The principal fraction of 16 N and 16 N nuclei is produced in the plasma facing components of
the divertor: in the inner vertical target and dome supports, where the water spends the most
of time at its passage in the toroidal channels, and in the cassette body.
The 16 N-nuclear density 6.9 10 10 cm -3 is several times higher than in the blanket cooling
water, and the 17 N-nuclear density is practically the same.
3.8.5 Total 16 N Gamma-Ray Source in the Outlet Pipes inside the Cryostat
The total 16 N gamma-source power of ~ 9 kW was calculated in 144 water pipes after their
exit from the upper ports (average pipe length ~ 11.4 m, water velocity ~ 5 m/s). The fraction
of gamma-ray energy released inside the cryostat is about 75% of the total gamma-ray source
in the pipes. The remainder of the energy is taken away by the water and is released outside
the biological shield.
The estimated 16 N gamma-ray heat source in the divertor coolant outside the divertor ports
(54 return pipes, 65 mm i.d., length ~1.7 m) is approximately 2 kW.
3.8.6 Absorbed Dose Rates and Coolant Pipe Activation
The water radiation conditions in the divertor coolant pipe environment were evaluated at the
cryostat and in a maintenance cell behind the biological shield using 2-D methods (See
reference 1 ).
The total 16 N decay power in three outlet pipes inside a pumping port is ~ 400 W. It is much
smaller than the total heat power ~ 9.7 MW removed from three divertor cassettes in a port.
The 16 N, 17 N densities, given in Table 3.8-5, attenuate along the ~5.8-m pipe from the
cassette body to the divertor port door by ~10%, 15%, respectively (water velocity ~ 6 m/s).
The rest of the pipes (~ 1.5 m) are located outside the port plugs, penetrate the cryostat where
they are practically unshielded. The pipes then pass through the biological shield into the
maintenance cell.
The next fluxes and dose rates arrived from the 16 N and 17 N -decay were estimated at a return
pipe surface:
16 N-decay gamma-flux 2.1 10 10 cm -2 s -1 ;
16 N-decay gamma-ray energy deposition 2.6 mW/cm 3 (SS);
Absorbed dose rate from 16 N-decay 1.2 10 3 Gy/h;
17 N-decay fast neutron flux 6 10 6 cm -2 s -1 .
These values decrease rapidly by 1-2 orders of magnitude at a distance of ~0.5-1 m from the
pipe.
1 V. Khripunov, R. T. Santoro, H. Iida, R. Plenteda, Mid-Plane Port Coolant Pipe Activation. NAG-78-03-03-
98, ITER Garching JWS, March 1998.
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ITER G 73 DDD 2 01-06-06 W0.1
The absolute fast neutron flux level caused by 17 N -decay neutron source in front of bioshield
is ~2-3 orders of magnitude lower than the main “background” flux level inside the
cryostat. At the same time the 16 N -decay photon flux is comparable or even higher than the
“background” photon flux.
The estimated equivalent dose rates caused by 16 N -decay photons behind the biological
shield are ~10-70 Sv/h, that excludes the presence of personnel in the maintenance cells near
the outlet water coolant pipe during reactor operation. After a few minutes, it decreases down
to the level of natural background due to its very short (7.13 sec) half life.
The contact equivalent dose rate ~ 60 μSv/h initiated by the internal 17 N -decay neutrons is
expected at the outlet “hot” pipe surface near the cryostat and in a maintenance cell. It is
estimated for the maintenance period one hour after shut down at the end of the DT-operation
phase and corresponds to 0.10 wt% Co in the pipe steel. It decreases down to ~ 6 μSv/h two
weeks after shutdown. This dose level is ~5-10 times lower than that produced in both inlet
and outlet cooling pipes from corrosion products and from the outer “background” neutron
source at the cryostat, respectively.
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ITER G 73 DDD 2 01-06-06 W0.1
4 Detailed Multi-Dimensional Analyses
4.1 Neutron Wall Loading Distribution on the First Wall
The neutron wall loading (the 14.1-MeV neutron current) is a key parameter for the
component materials selection of a fusion reactor. It defines design, safety and ultimately
nuclear performance of a system. The poloidal distribution on the first wall is usually used in
many cases in the design process as a normalisation factor for a quick estimation of the
poloidal nuclear respond variations and their local values basing on the results calculated for
average conditions. The distribution provides the peak values which are used to size the
shield, determine components’ lifetimes, estimate damage levels, and assess the radiation
environment around the tours. Further more, and accurate estimate for the wall loading at
some critical shielding regions (in inboard part of the machine and behind the divertor) is
essential to warrant adequate protection for the inner legs of the TF coils.
The poloidal distribution of the neutron wall loading was investigated in references 1 and 2 ,
using 3-D ray tracing Monte Carlo codes. For that purpose the geometrical description of the
first wall surface formed by the plasma facing structures and incident 14.1-MeV neutron
source configurations in the plasma chamber were modelled in details.
4.1.1 Neutron Source Distributions
Different kinds of plasma density and neutron source models were used for that. Originally, a
simplified DT-neutron distribution in plasma was implemented in the general ITER models 1
using the previous neutron source model developed for the ITER-98 and referred to the
current first wall configuration and the fusion power 500 MW. The source region was divided
into five layers (cells) with the scrape-off layer separating them from the first wall (Figure
4.1-1).
Figure 4.1-1 Plasma Region approximated by Five Cells with a Uniform Source in
each Layer
1 D. Valenza, H. Iida, R. Plenteda, Three- Dimensional Nuclear Analysis of the First Wall of the RC-ITER
Reactor (IAM option), G 73 RI 104 99-04-14 W 0.1 (NAG-129-04-14-99). Garching, April 1999.
2 G. Ruvutuso and H. Iida, Neutron Wall Loading with a Non-inductive Operation Plasma. NAG-156-23-05-
00, Garching, May 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
DT-source neutron parameters (co-ordinates, direction cosines for isotropic angular
distribution and a neutron energy from the Gaussian fusion spectrum) were sampled by
determining the plasma region in which the neutrons are born uniformly in one of the five
source layers. The probabilities assigned to these layers are correspond to the DT-reaction
rate, varying from 0.5 to 0.01 in the inner, middle and outer cells, respectively.
A well known analytic description of a 14.1 MeV neutron source intensity IDT-n,
IDT-n ~ n 2 T 2 (1- r 2 /rs 2 ) P
was also used to investigate a sensitivity of the neutron wall load distribution to a fusion
reaction rate profile. The next notations are used for this description:
n - plasma density,
T - plasma temperature,
r - current co-ordinate relating to the source center,
rs - plasma surface,
P = 3-6, depending on the operation regime at different Q-factors 1 .
A modelling shows a week dependence of the neutron wall loading distribution on the Pfactor.
The local values differ by ±1-3% for different P-values. Even in the case of a
collapsed “stringwise” neutron source (rs ~ 0.2-0.3 m instead of a more realistic rs ~ 2-3.4 m)
the difference in the results does not exceed ±3-8% locally.
Then, a map of the neutron source distribution, corresponding to the actual DT-plasma
density profile in the referent inductive operation regime I (Pfus=500 MW, Q=10, Ip=15 MA
operation with heating during current ramp-up), was developed in accordance with reference
2 for the general ITER model.
The pointwise source in plasma region (Figure 4.1-2) is represented by a single 40 x 40
matrix and each cell (10.8 cm radial x 20.8 cm vertical) is assigned a DT-neutron born
probability. In comparison with a more flat distribution determined for the ITER-98 plasma,
the current distribution drops sharply from the center to the outer regions of the plasma.
The neutron source intensity in Figure 4.1-2 varies as a function of R-Z location. Cells are
sampled randomly to determine the source particle origin and are assigned direction cosines
and constant statistical weight. A Gaussian energy distributed DT neutrons peaked around
14.1 MeV was used for source particle energy sampling. Cells outside the first wall envelope
have zero probability.
Source particles travel through the void in the plasma chamber until they cross the wall. In
the DT-neutron wall loading calculations, particles are killed upon crossing the first wall
surface. Surface current tallies represent the neutron wall loading (Figure 4.1-3).
1 V. Chuyanov, private communication. ITER Garching JWS, December 1999.
2 Yoshiki Murakami, Private Communication, ITER Naka JWS), 15 May 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 4.1-2 DT-Neutron Source Strength R, Z-Distribution
in the Plasma Chamber
5 m
0.59
MW/m 2
0
-5 m
5 m
0.43
MW/m 2
~ 0.55
MW/m 2
0.78
MW/m 2
Figure 4.1-3 Neutron Wall Loading Distribution on the ITER First Wall
(fusion power 500 MW)
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ITER G 73 DDD 2 01-06-06 W0.1
An analysis 2 indicates a relative error in the neutron wall loading introduced by
approximating the source profile by the five-layer-distribution in comparison with the
pointwise distribution. The largest difference between the calculated maximum neutron wall
loading values refers to the central first wall segments (an underestimation ~ 5% in inboard
module 3, and an overestimation ~9% in the outboard module 13). This is mainly due to the
assuming plasma core location slightly shifted in the upper direction in comparison with a
realistic plasma center position.
4.1.2 Averaged Values and Local Maximums
Table 4.1-1 gives the peak and average neutron wall loading values in different regions
calculated for two neutron source distributions: (i) the nominal inductive operation (Pfus =
500 MW) and (ii) the non-inductive operation regime I (DRG scenario 4 (See reference 1 ) at
the lower fusion power (Pfus= 400 MW), smaller minor radius (a= 1.85 m instead of 2 m) and
outward shifted plasma core (R= 6.35 m instead of 6.2 m).
In the case of the hybrid non-inductive operation the relative peak value at the outboard
equatorial plane is slightly, by ~3% higher than during the reference inductive operation (See
Table 4.1-1).
Table 4.1-1 Average and Peak Neutron Wall Loading Values
for the Nominal Inductive Operation I and the Non-Inductive Operation I
Surface area, m 2
Average, MW/m 2
Peak, MW/m 2
Average, MW/m 2
Peak, MW/m 2
Inboard
FW
Outboard
FW
Total
FW
Divertor
Entrance
Total
~200 ~470 670 ~60 ~730
Inductive Operation I (Pfus= 500 MW, R= 6.2 m, a= 2 m)
0.43 0.62 0.57 0.47 0.56
0.62 0.76 0.49
Non-Inductive Operation I (Pfus= 400 MW, R= 6.35 m, a= 1.85 m)
0.32 0.50 0.45 0.37 0.44
0.46 0.62 0.38
However, the fusion power is by 20% lower than in case of the reference operation regime.
The maximum inboard and outboard neutron wall loads are ~0.46 and ~0.62 MW/m 2 ,
respectively, these are also by ~20% lower than in the reference case.
1 Design Requirements and Guidelines Level 1 (DRG1). G A0 GDRD 2 00-12-01 W 0.5. December 1, 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
Thus the typical average neutron wall loading including the divertor entrance surface is ~
0.55 MW/m 2 for both regimes at the same nominal fusion power of 500 MW, the inboard
peak is 0.60 MW/m 2 , and the outboard peak ~0.78 MW/m 2 .
Using the average neutron wall loading values along with the areas of the different regions,
the fractions of source neutrons impinging directly on the inboard, outboard and divertor
regions of ITER can be determined as 21%, 72% and 7%, respectively.
4.1.3 Flux-to-Current Ratios
In some cases the neutron wall loading as a normalisation factor is not valid for nuclear
respond estimates. For example, nuclear responses in the inboard part of the first wall and
blanket, evaluated on the base of preliminary accurate calculations for the central outboard
part of the machine, are usually underestimated by 30-60% or even more in the top and
bottom regions and in the divertor targets.
The reason is that in fact they are functions of neutron fluxes, while the neutron wall loading
is the incident 14.1 MeV neutron current. At the same time, the using of the uncollided 14.1
MeV neutron flux instead of the neutron wall loading can diminish this uncertainty in
response estimates to about 10% locally.
In order to get detailed poloidal distributions of the neutron wall loading and the 14.1 neutron
flux, the actual geometrical configuration of the first wall formed by the front surfaces of the
blanket modules was segmented. Both the 14.1 MeV neutron current crossing the first wall
surface and the uncollided neutron flux were tallied simultaneously in the set of detectors
shown in Figure 4.1-4.
Z [cm]
500
400
300
200
100
-100
-200
-300
-400
0
0 100 200 300 400 500 600 700 800 900 R [cm]
Figure 4.1-4 First Wall Surface Profile and Tallies
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 4.1-5 shows the poloidal variations of the neutron wall loading [MW/m 2 ] and the
uncollided 14.1 MeV neutron flux [10 14 cm -2 s -1 ] in the inboard and outboard regions as a
functions of the poloidal length measured in the clockwise direction from the lower corner of
the inboard first wall.
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0 200 400 600 800 1000 1200 1400 1600 1800 2000
0.50
0.45
0.40
0.35
0.30
0.25
0.20
0.15
0.10
0.05
#1 #2 #3 #4 #5 #6 #7 #8 #9 #10 #11 #12 #13 #14 #15 #16 #17 Divertor
poloidal length (cm)
n. wall loading 14 MeV [MW/sqm], R=6.2, a=2
(Pf=500 MW)
n. wall loading 14 MeV [MW/sqm], R=6.35, a=1.85
(Pf=500 MW)
#1 #2 #3 #4 #5 #6 #7 #8 #9 #10 #11 #12 #13 #14 #15 #16 #17 Divertor
0.00
0 200 400 600 800 1000 1200 1400 1600 1800 2000
poloidal length (cm)
14 MeV Flux [ 1E+14 n / (sqcm*s) ], R=6.2, a=2
(Pf=500 MW)
14 MeV Flux [ 1E+14 n / (sqcm*s) ], R=6.35, a=1.85
(Pf=500 MW)
Figure 4.1-5 Neutron Wall Loading and Uncollided Flux versus Poloidal Length
(for inductive and non-inductive operation regimes)
The results shown in Figure 4.1-5 for the inductive and non-inductive operation regimes are
referred to the nominal fusion power of 500 MW. The fractional standard deviations
associated with these results received by the Monte Carlo method is ~0.2 %.
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ITER G 73 DDD 2 01-06-06 W0.1
A relation between the DT neutron wall loading and the uncollided DT neutron flux at the
first wall surface characterises the mean incidence angle of the direct neutron current through
the first wall 1 (in fact their cosines).
The flux-to-current ratio (F/J) is shown in Figure 4.1-6 as a function of the poloidal length.
The corresponding function of the incident neutron angular distribution at the first wall is
given in Figure 4.1-7 (reference 2 ).
Φ / j
2.5
2.0
1.5
1.0
0 200 400 600 800 1000 1200 1400 1600 1800 2000
poloidal length (cm)
Figure 4.1-6 Uncollided Flux-to-Current Ratio (F/J) versus Poloidal Length
degrees
90
80
70
60
50
40
30
20
10
0
0 200 400 600 800 1000 1200 1400 1600 1800 2000
poloidal length (cm)
Figure 4.1-7 Average Angle between a Surface Normal
1 U. Fischer, Qualification of Neutronic Blanket and Shielding Calculations in a One-Dimensional Approach to
a Tokamak Reactor. Fusion Technology, Vol. 22, pp 251-270 (1992).
2 G. Ruvutuso and H. Iida, Three-Dimensional Nuclear Analysis of the First Wall of ITER-FEAT (MCNP 3D
model ver. 1). NAG-149-03-03-00, Garching JWS, March 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
and the Incident Particle Trajectories versus Poloidal Length
Numerically, the F/J ratio at the first wall surface varies from 1.8 at the bottom inboard
modules No. 1, 2, to 1.6 at modules No 3-5, and again to 2.1 at the upper modules No 6. At
the outboard modules this ratio is lower ~1.3-1.5 and at the divertor dome surface ~1.45.
This variation explains a possible uncertainty in estimates introduced by using the neutron
wall loading as a normalisation factor.
4.2 Blanket
4.2.1 Heat Deposition in the Blanket Modules
In the basic 3D model a careful description of the segmentation of the inner vacuum vessel
components has been made (see Figure 4.2-1). There are 17 blanket modules poloidally, the
inboard modules being numbered from one to eight, the remaining belonging to the outboard
part. The toroidal segmentation is different in the inboard and outboard part: per 20° sector
there are 2 outboard modules and 1 inboard.
The overall radial thickness of each blanket module is 45 cm. The radial layout of the
blanket model is a layered structure, with the front beryllium armour (1 cm thick) followed
by a 2 cm thick heat sink. The remaining 42 cm is the bulk shield. It has been represented as
a homogenised mixture of 84% steel and 16% water.
Filler shield elements are inserted in toroidal gaps between the blanket modules, with an
assumed homogenised material composition of 50% steel and 50% water. The poloidal layout
of manifolds has also been described.
The calculated nuclear heating in the blanket system components is summarised in Table 4.2-
1 and Figure 4.2-2
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 4.2-1 Picture of the blanket Module
Table 4.2-1 Nuclear Heating in the Blanket Modules
Inboard first wall 30
Inboard blanket (w/o first wall) 104
Outboard first wall 59
Outboard blanket (w/o first wall) 230
Filler wedge elements 7
Manifolds 4
Equatorial port plug 58
Upper port plug 9
Total 502
(MW)
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ITER G 73 DDD 2 01-06-06 W0.1
7
1
17
Component,
Module #
Nuclear
Heating* in
the whole
modules
Nuclear
Heating* in
the first
wall only
1 890 kW 192 kW
2 982 219
3 1,172 250
4 1,110 241
5 871 202
6 683 168
7 822 185
8 914 198
9 828 174
10 2,005 424
11 1,793 367
12 2,027 407
13 1,013 202
14 1,349 268
15 2,492 494
16 2,455 500
17 2,082 424
Manifolds 211
Fillers 411 -
Equatorial
Plug
3,261 -
Upper Plug 500 -
* values are for a single 20˚ sector (kW) Total 27,870 4,915
Figure 4.2-2 Nuclear heating in the blanket modules
4.2.2 Flexible joint analysis
Two issues are aroused concerning the flexible joints which mechanically support blanket
modules from the vacuum vessel. One is the radiation damage of the flexible joints,
especially that of Ti alloy and Incoloy bolts. Another is the effect of void region included in
the joints on the nuclear heating in the TF coil.
4.2.2.1 Model description
Figure 4.2-3 shows 3D view of the partial model and Figures 4.2-4 and 4.2-5&6 show detail
of the flexible joint geometry. The flexible joint is a cylinder with 13 cm of diameter and 23
cm of length and placed in the hole with 15 cm diameter. The largest void space in the joints
is roughly cylindrical geometry with the 11 cm of diameter and 9 cm of height.
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ITER G 73 DDD 2 01-06-06 W0.1
TF Coil
Plasma
Outborad
Reflector
Blanket
Vacuum
Vessel
Figure 4.2-3 Partial 3D model for the flexible joint analysis
15 cm
9 cm
23 cm
Figure 4.2-4 Detail of the flexible joint
Figure 4.2-5 The geometry of the flexible joint
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 4.2.6 The detail geometry of the flexible joint
4.2.2.2 Radiation Damages of the Flexible Cassette (Ti alloy) and a Bolt (Incoloy) 1
Results of radiation damage calculation with the 3-D model are shown in Figure 4.2.7 for the
flexible cassette and Figure 4.2-8 for the Inconel bolt. The maximum damage in the flexible
cassette (Ti alloy) is ~0.014 dpa and is low enough causing no significant problem.
The maximum damage in the Inconel bolt end of the blanket attachment system is ~ 0.021
dpa. This damage is low and will probably not result in significant property changes of the
structure materials.
All the above values are normalised to the local peaking fluence 0.42 MWa/m 2 at outboard
first wall, which is consistent with the average neutron fluence of 0.3 MWa/m 2
1 S. Sato, Y. Ohara and M. Akiba, “Radiation streaming analysis through equatorial port”, JP HT report of
D469-JA, 2001, June
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ITER G 73 DDD 2 01-06-06 W0.1
2.9x10 -3
Cartrige (Ti-6Al-4V)
(0.0328)
1.7x10 -3
Void
(0.0353)
DSCu
5.8x10 -3
(0.0306)
Vacuum Vessel Blanket
Bolt (INCONEL718)
1.4x10 -2
(0.0265)
5.0x10 -2
(0.0717)
Void (φ=10mm)
0.13
(0.0542)
(unit : dpa)
0.99
(0.0450)
0.37
(0.0517)
First Wall
2.9
(0.0323)
Front access hole (φ=30 mm)
Figure 4.2-7 Neutron damage of flexible support cartrige (Ti-6Al-4V) at inboard
mid plane blanket under average neutron fluence of 0.3MWa/m 2
Cartrige (Ti-6Al-4V) Vacuum Vessel Blanket
6.5x10 -3
(0.0338)
4.3x10 -3
Void
(0.0320)
9.3x10 -3
(0.0328)
1.2x10 -2
(0.0385)
1.4x10 -2
(0.0430)
2.1x10 -2
(0.0520)
4.7x10 -2
(0.0702)
0.12
(0.0535)
(unit : dpa)
0.96
(0.0453)
0.37
(0.0542)
First Wall
2.9
(0.0319)
DSCu Void (φ=10mm)
Bolt (INCONEL718)
Front access hole (φ=30 mm)
Figure 4.2-8 Neutron damage of flexible support bolt (INCONEL 718) at inboard
mid plane blanket under average neutron fluence of 0.3MWa/m 2
4.2.2.3 The Effect of Flexible Joint Module on Nuclear Heating in the TF Coil
inboard Legs 1
Existence of flexible joints in the vacuum vessel increases nuclear heating in TF coils. Its
effect is especially important for evaluating the heating in the TF coil inboard legs. Since it is
not practical to include the geometry of the flexible joints in the basic 3-D ITER model 2 , the
effect has to be assessed by the separate analysis which uses a simpler “partial model” of
ITER.
1 S. Sato and H. Iida, NAG-169 “The Effect of Flexible Joint Module on Nuclear Heating in the TF Coil inboard
Leg” Nov. 2000
2 G. Ruvutuso,H. Iida and L. Petrizzi, NAG-168, Updated of Basic 3-D model of ITER for Monte Carlo nuclear
analyses with MCNP
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ITER G 73 DDD 2 01-06-06 W0.1
Calculation results are shown in Table 4.2-2. The ratio of nuclear heating in the TF coil in the
cases of with and without the flexible joints is 1.20. In the case with flexible joints, the
heating due to the particle passing through the void space in the flexible is also calculated by
flagging technique. Its contribution is about 26 - 28% of the total heating suggesting that
increase of heating is surely caused by the existence of the void. The difference between the
values of 1.2 and 1.26 – 1.28 can be justified when we think about the fact that the “void
space” should have a few % of contribution even when the space is filled with material.
Table 4.2-2 TF coil inboard leg nuclear heating with and without flexible joints
TF coil Nuclear Heat (1) w Flex (2) w/o Flex Ratio (1)/(2)
Total (W) 42.3 (0.023) 35.3 (0.028) 1.20
Flagged (W) 9.3 (0.030)
Total/ (Total - Flagged) 1.28
(Total + Flagged)/Total
1.26
4.2.3 Helium production in the branch pipe and at the surface of the vacuum
vessel 1
Helium production in stainless steel is an important parameter for those parts of the vacuum
vessel and blanket that need to be re-welded during maintenance or replacement. Heproduction
rates in the cooling branch pipes of the blanket modules have been estimated in
detail with a 3-D model shown in Figures 4.2-9 and 4.2-10. The blanch pipe and surrounding
‘grooved area’ configuration is simulated in detail. An average first wall neutron fluence of
0.3 MWa/ m 2 (0.42 MWa/m2 at outboard maximum) is assumed in the analysis.
Figure 4.2-11 shows the helium production in the front access hole and branch pipe of the
blanket module obtained with the 3-D model for the reference structural material. The
reference material, SS316L(N)-IG has low boron content (~ 10 wppm ). Boron in the steel
gives large contribution in helium production through its huge B 10 (n,α) reaction cross section
in the region where thermal neutron flux dominates. The helium production at the welding
part of the branch pipe is estimated to be 1.2 appm.
1 S. Sato and H. Iida, NAG-169 “The Effect of Flexible Joint Module on Nuclear Heating in the TF Coil inboard
Leg” Nov. 2000
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ITER G 73 DDD 2 01-06-06 W0.1
Front access hole (φ=30mm)
Blanket
A
Vacuum vessel
A
First wall
Void
Coolig water pipe (ID: 42.6 mm,
thickness: 3 mm)
Figure 4.2-9 Horizontal cross section of a blanket module and a branch pipe
Grooved
Area
A-A B-B
Figure 4.2-10 Vertical cross section of a blanket module and a branch pipe
unit: appm
(fsd: %)
72(5.3)
17(4.2)
9.1(8.6)
6.3(5.3)
3.6(4.5)
1.9(2.1)
1.1(1.8)
1.2(1.4)
1.2(1.2)
0.65(0.73) 0.85(0.81)
0.48(0.99)
Figure 4.2-11 Helium production distribution in SS with 10 wppm B along the front
access hole and the cooling water branch pipe (Neutron fluence: 0.3 MWa/m 2 )
Nuclear Analysis Report Page 61
B
1.1.1.1.1.1.1.1.1 B

ITER G 73 DDD 2 01-06-06 W0.1
Design limit of helium production for re-welding part is 1 appm for thicker welding. For a
thinner welding, like that of branch pipe, the limiting value 3 appm (see Chater 2) and rewelding
of the branch pipe is possible. Helium production in the steel with higher boron
content (20 wppm), similar to usual stainless steel, is shown in Figure 4.2-12. The
production become larger by nearly twice and can become problematic for re-welding when
we think about uncertainty included in the analysis (see Chapter 9).
unit: appm
(fsd: %)
93(5.3)
23(4.2)
13(8.6)
9.0(5.3)
5.4(4.5)
2.9(2.1)
1.7(1.8)
2.2(1.4)
2.1(1.2)
1.2(0.73) 1.6(0.81)
0.89(0.99)
Figure 4.2-13 Helium production distribution in SS with 20 wppm B along the front
access hole and the cooling water branch pipe (Neutron fluence: 0.3 MWa/m 2 )
Toroially continuous gaps between blanket modules cross the field joint line of the vacuum
vessel. Thus it is important to estimate the helium production in that part taking account of
the streaming effect through the gap. Helium production at the surface of the vacuum vessel
is calculated with the same model, giving the value for the inboard local peak as shown in
Fig. 4.2-14.
Nuclear Analysis Report Page 62

ITER G 73 DDD 2 01-06-06 W0.1
unit: appm (fsd)
TF coil
Vacuum vessel
2.84E-01(10.72%)
Flexible support
4.40E-01(8.79%)
Nuclear Analysis Report Page 63
Gap
Blanket
Figure 4.2-14 Helium production in SS of vacuum vessel under the gap between
adjacent blnaket modules. (Average neutron fluence: 0.3 MWa/m 2 corresponding to
0.42 MWa/m 2 at maximum on the outboard)
The value is 0.44 appm which is a little smaller than that estimated in the section 3.3 ( 0.6
appm) and has an enough margin (see section 9.4) than design limit (1 appm).
4.3 Vacuum Vessel
4.3.1 Integral nuclear heat 1
The vacuum vessel has been modeled according to its layout by three layers. There are two
robust outer shells 6 cm thick both made of pure SS 316 L(N) IG. The thickness of the filler
region between the two shells, varies along the poloidal direction and has been described as
an homogenized material mixture of borated steel 60% (2% in weight of natural boron) and
40% water. The overall thickness of the vacuum vessel at the equator is 33.7 cm in the
inboard and 75 cm in the outboard.
1 G. Ruvutuso, H. Iida and L. Petrizzi,”Nuclear Heat deposition in the Blanket and Vacuum Vessel
by Monte Carlo nuclear analyses with MCNP”, NAG-171-15-11-00, November 2000

ITER G 73 DDD 2 01-06-06 W0.1
Figure 4.3.1-1 3D Picture of the Vacuum Vessel Model out of the other machine
components
Results of analysis are summarized in Table 4.3.1-1. The heating on the port walls can be
dependent on the port plug configuration, but the contribution of the port walls to the total
is marginal.
Table 4.3.1-1 Nuclear Heat Deposition in the Vacuum Vessel
Segment # kW
Inboard Vacuum Vessel C1 – C4 997
Top Vacuum Vessel C5 – C9 819
Outboard Vacuum Vessel C10 – C14 3260
Bottom Vacuum Vessel C15 – C19 324
Blanket Support 1296
Port Walls 396
Total 7092
4.3.2 Local nuclear heat
The vacuum vessel model has been divided in 19 poloidal segments not with the same
poloidal length (see Fig 4.3.2-1) and in 3 segments in toroidal direction (see Fig. 4.3.2-2),
Nuclear Analysis Report Page 64

ITER G 73 DDD 2 01-06-06 W0.1
In table 4.3.2-1 the total power generated in each poloidal segment is given. The numbers
can vary with the poloidal extension, the wall loading behavior and also with the toroidal
extension. In fact from the table, C-12 looks less heated than C11 or C-13. This is because
C-12 is in toroidal direction interrupted by the port entrance (see Fig. 4.3.2-2),
Table 4.3.2-1 Total nuclear power in the vacuum
vessel for each polidal segment
Vacuum Vessel Total nuclear
Poloidal Segment heating (kW)
C-1 92
C-2 328
C-3 360
C-4 214
C-5 106
C-6 175
C-7 297
C-8 27
C-9 214
C-10 704
C-11 954
C-12 626
C-13 828
C-14 15
C-15 9
C-16 122
C-17 20
C-18 112
C-19 63
total (MW) 5.4
200.0 400.0 600.0 800.0 1000.0
Nuclear Analysis Report Page 65
C-5
C-3
C-1
C-19
C-7
0 cm
C-9
C-15
C-11
C-13
Figure
C-17
4.3.2-1 Vacuum vessel
poloidal segmentation, with labels.
In Figures 4.3.2-3 and 4.3.2-4 the power density generated in the inner 6 cm steel layer is
shown for the central toroidal segment and for the lateral segments respectively, in function
of the poloidal length. There are some difference in the nuclear heating between the two
toroidal sectors: the central one and the lateral one. This is due to the influence of the
poloidal gaps dividing the blanket modules in front of the vacuum vessel and in the equatorial
port (Figures 4.3.2-5 and 4.3.2-6).

ITER G 73 DDD 2 01-06-06 W0.1
0.14
W/cm 3
0.12
0.10
0.08
0.06
0.04
0.02
0.00
Thermal shield,
Inboard
manifol
d
Inboard Blanket
Module
Vacuum Vessel
Lateral Cells
Vacuum Vessel
Central Cell
Outboard
manifold
Outboard
Blanket
Module
Figure 4.3.2-2 Equatorial cross section of the 3D model
C-1 C-2 C-3 C-4 C-5 C-6 C-7
C-8
C-9 C-10 C-11 C-12 C-13
C-15
C-14 C-16
Eq. Plug
Eq. Plug
C-17
C-18
0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600
poloidal length [cm]
Figure 4.3.2-3 Poloidal distribution of power density of the inner shell for the central
toroidal segment of vacuum vessel (6 degrees wide)
Nuclear Analysis Report Page 66
C-19

ITER G 73 DDD 2 01-06-06 W0.1
0.14
W/cm 3
0.12
0.10
0.08
0.06
0.04
0.02
0.00
C-1 C-2 C-3 C-4 C-5 C-6 C-7
C-8
C-9 C-10 C-11 C-12 C-13
C-15
C-14 C-16
C-17
C-18
0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600
poloidal length [cm]
Figure 4.3.2-4 Poloidal distribution of power density of the inner shell for the two
lateral segments of vacuum vessel (7 degrees wide).
1.E+00
W/cm 3
1.E-01
1.E-02
1.E-03
0 5 10 15 20 25 30 35
Radial distance
cent. Segm..
lat. Segm.
Figure 4.3.2-5 Radial distribution of the nuclear heating at the inboard mid-plane
(C-3) of the Vacuum Vessel
Nuclear Analysis Report Page 67
C-19

ITER G 73 DDD 2 01-06-06 W0.1
1.E+00
W/cm 3
1.E-01
1.E-02
1.E-03
1.E-04
1.E-05
1.E-06
0 10 20 30 40 50 60 70 80
Radial distance
cent. Segm..
lat. Segm.
Figure 4.3.2-6 Radial distribution of the nuclear heating at the outboard mid-plane
(C-12) of the Vacuum Vessel
The gaps, which run poloidally and toroidally dividing the blanket modules, produce peak
values of the heating at the inner layer of the vacuum vessel. The two gaps are both 20 mm
wide. There are peaks in toroidal direction due to the presence of poloidal gaps, and in
poloidal direction due to the presence of toroidal gaps. The peak value is defined as the ratio
of the maximum value over the average one. At the intersection of the two gaps, a larger peak
is expected. Figures 4.2.2-7 and 4.2.2-8 give different sections of the inboard part of the
machine. More is given in the reference 1 .
O
U
T
E
R
S
H
E
L
L
V
A
C
U
U
M
V
E
S
S
E
L
I
N
N
E
R
S
H
E
L
L
Blanket
Module no. 4
Blanket
Module no. 3
Figure 4.3.2-7 Poloidal section along A-A through the manifolds that run behind
the blanket modules.
1 G. Ruvutuso, H. Ida, L. Petrizzi, NAG 171-15-110, “Nuclear Heat Deposition in the Blanket and in the
Vacuum Vessel”, Nov. 2000.
Nuclear Analysis Report Page 68
A
Toroidal
gap
A
Equatorial
section
Poloidal
gap

ITER G 73 DDD 2 01-06-06 W0.1
TFC
T
H
E
R
M
A
L
S
H
I
E
L
D
Gap
between
modules
no. 4
Gap
between
modules
no. 3
B B
Figure 4.3.2-8 Poloidal cross section along B-B through the gaps between the
modules
In Table 4.3.2-2 the heating values are shown taken at four poloidal locations: (C-3, C-7, C-
12 and C-18, see Table 4.3.2-1). In the first line there are the values averaged over the 60 mm
thickness of the inner steel layer. In the second line there are the values averaged over the
first 10 mm thickness. Then peaking factors along the toroidal and poloidal directions are
given due to the poloidal and toroidal gaps. The peak value at the intersection is given as
well. Finally, the latest value in the last row gives the ratio between the heating value at the
intersection, and the lowest heating value behind the blanket. The latter is close to the value
that would result from bulk shielding conditions, similar to a 1-D model, in which no gaps
are considered.
Table 4.3.2-2 Values on the wall of the inner shell in a few sectors and peak values
in the cross point.
Poloidal Sector C-3 C-7 C-12 C-18
Mean value (W/cm 3 ) 0.09 0.07 0.13 0.02
Wall value (W/cm 3 ) 0.13 0.10 0.19 0.03
Peak
factors
Toroid./Pol.
Intersection
1.7/2.8
3.5
1.7/2.8
3.5
1.3/2.8
3.2
-
Peak factor for 1D 4.6 4.6 4.2 -
Equatorial
section
Nuclear Analysis Report Page 69

ITER G 73 DDD 2 01-06-06 W0.1
4.3.3 Gap streaming onto the vacuum vessel 1 , 2
Enhanced thermal stresses in the vacuum vessel due to the nuclear heat peaking is one of the
issues for the vacuum vessel design. It is in part caused by the radiation streaming through
the gaps between blanket modules. A Monte Carlo transport analysis was conducted to give
detailed nuclear heat distribution on the front surface of the vessel. The results were used as
input data for the consequent thermal analysis by the designer.
4.3.3.1 Calculation model
Figure 4.3.3-1 and 4.3.3-2 show a horizontal and a vertical view of the model. Inboard and
out board blankets are annuluses (by means of mirror reflection boundaries on the both sides
of sectors) and vacuum vessel is modelled with a slab. In the direction of the cylinder axis, an
upper and a lower reflectors (reflecting material) were placed rather than mirror reflection
boundaries in order to take account of the finite height of the real plasma.
The analysis is conducted for three cases; namely no manifold in the V-shape gap ( case 1:
Figure 4.3.3-3), 5 manifolds (case 2: Figure 4.3.3-4) and two manifolds (case 3: Figure4.3.3-
5). The second corresponds to the vacuum vessel behind # 11,12 blanket modules and the last
behind #16 module. They are shown in Figures 4.3.3-3 and 4.3.3-4.
Plasma
Blanket
Gap
Figure 4.3.3-1 Horizontal view of the MCNP model
Vacuum
Vessel
1 H. Iida and L. Petrizzi , “Nuclear Heat Peaking on the Vacuum Vessel Front Shell due to V-shape Gaps
between Blanket Modules”, NAG-140-11-22-99, December, 1999
2 H. Iida , ”Nuclear Heat on the Vacuum Vessel in case of V-shape Gaps with Manifolds between Blanket
Modules”, NAG-162, July 2000
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ITER G 73 DDD 2 01-06-06 W0.1
Inboard
Reflector
Plasma
Upper Reflector
Lower Reflector
Vacuum
Vessel
Outboard
Blanket
Figure 4.3.3-2 Vertical view of the MCNP
model
Figure 4.3.3-4 Zoom up of the Vshaped
gap geometry with 5 manifolds
(case 2)
4.2.3.2 Results of Analysis
Nuclear Analysis Report Page 71
9.1 cm
2 cm
Figure 4.3.3-3 Zoom up of the Vshaped
gap geometry (case 1)
Figure 4.3.3- 5 Zoom up of the Vshaped
gap geometry with 2 manifolds
(case 3)
4.2.3.2.1 Nuclear Heat on the Vacuum Vessel without Manifolds between Blanket
Modules (case 1)
Neutron flux distributions obtained by MCNP (3-D Fast and Total) are compared with the
results by 1-D code ANISN as shown in Figure 4.3.3-6. A 14 MeV neutron current at the
outboard first wall is normalised to be 0.64 MW/m2 (average) for the both cases. The 1-D
calculation naturally corresponds to the case of “no gap”. The comparison implies that the

ITER G 73 DDD 2 01-06-06 W0.1
peaking factor at the vacuum vessel surface be a little less than factor ten, which is consistent
with the preliminary prediction by the handy method 1 (see Figure 4.3.3-7 ).
11
10
9
8
7
6
5
4
3
2
1
1.E+15
1.E+14
1.E+13
1.E+12
1-D Fast
1-D Total
3-D Fast
3-D Total
1.E+11
800 820 840 860 880 900
Distance from the Torus Axis (cm)
Nuclear Analysis Report Page 72
Blkt
Figure 4.3.3-6 Comparison of neutron fluxes between MCNP and ANISN
0
0 1 2 3 4 5 6
Gap Width (cm)
Effective
Width:5.5 cm
Figure 4.3.3-7 Peaking factor predicted
by the handy method
Nuclear Heating (W/cc)
1.E+01
1.E+00
1.E-01
VV
Blanket
MCNP
VV Shell
1.E-02
820 840 860 880 900
Distance from the Torus Axis (cm)
Figure 4.3.3-8 Nuclear heat distribution
on the outboard region by 1-D analysis
The peaking factor was predicted by the handy method 3 , which can handle only simple
geometry penetrations. When we take average of the gap width as effective gap width,
peaking factor of about 9 is obtained. Figure 4.3.3-8 shows nuclear heating distribution
obtained by the 1-D calculation with a value behind the bulk shield in 3-D calculation. All of
these are showing consistency between 1- and 3- D calculations.
1 H. Iida, R. T. Santoro, D. Valenza, and V. Khripunov, “A Handy Method For Estimating Radiation Streaming
Through Holes In Shield Assemblies”, Fusion Engineering and Design 43 (1998) 1-13.

ITER G 73 DDD 2 01-06-06 W0.1
The nuclear heat distributions along the surface of the vacuum vessel are shown in Figures
4.3.3-9 and 4.3.3-10. The value of heating rate is about 0.75 W/cc at the center of the gap
and decrease down to half at 7.5 cm from the center of the gap.
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
0 to 1 cm 1 to 2 cm
2 to 3 cm 3 to 4 cm
4 to 5 cm 5 to 6 cm
0 20 40 60 80 100
Distance from the center of the gap (cm)
Figure 4.3.3-9 Nuclear heat distribution in the direction parallel to the VV surface
1.00
0.10
0.01
0 cm 2 cm 4.5 cm 8 cm 12.5 cm
17.5 cm 25 cm 35 cm 45 cm 100 cm
0 1 2 3 4 5 6
Distance from the surface of the vacuum vessel (cm)
Figure 4.3.3-10 Nuclear heat distribution in the direction normal to the VV surface
Nuclear Analysis Report Page 73

ITER G 73 DDD 2 01-06-06 W0.1
Distribution of the nuclear heat integrated over the front shell thickness (6 cm) is shown in
Figure 4.3.3-11
3.5
3.0
2.5
2.0
1.5
1.0
0.5
0.0
0 20 40 60 80 100
Distance from the center of the gap (cm)
Figure 4.3.3-11 Nuclear heating distribution integrated over the vacuum vessel
thickness
The peaking factor in the nuclear heat density in the vacuum vessel front shell is about 8.6
[ratio of heat load at center (3.07W/cm 2 ) and that at 100 cm away from the
center(0.355W/cm 2 )], which is a little less than the preliminary estimate value of ~9 predicted
by the handy method. The peak value of nuclear heat integrated over the vacuum vessel front
shell is about 3 W/cm 2 , which may cause too high thermal stress in the vessel 1 .
4.3.3.2.2 Nuclear Heat on the Vacuum Vessel in case of V-shape Gaps with Manifolds
between Blanket Modules in the (case 2 and 3)
In this section, the nuclear heat distributions were calculated putting manifolds for blanket
cooling inside the V-shape gap. The manifolds were expected to reduce the nuclear heat
peaking then reducing the thermal stress.
The nuclear heat distributions are shown in Figures 4.3.3-12 and 4.3.3-13. The maximum
values of heating rate are about 0.34 W/cc with 5 manifold (case 2) and 0.6 W/cc with 2manifold
(case 3). Corresponding peaking factors are ~4 for the former and ~6.5 for the
latter.
According to the results of stress analysis 4 , a peaking factor of ~ 4 is acceptable but not that
of 6.5. In order to reduce this peaking factor down to around 4, dummy stainless shield plate
with 3 cm thickness will be required at front of (or back side of) the manifolds.
1 G Sannazzaro, IDoMS : G 15 MD 196 00-07-24 W0.1, " Thermal stress in the ITER VV outboard wall due to
nuclear heat load"
Nuclear Analysis Report Page 74

ITER G 73 DDD 2 01-06-06 W0.1
0.40
0.35
0.30
0.25
0.20
0.15
0.10
0.05
0 to 1 cm 1 to 2 cm
2 to 3 cm 3 to 4 cm
4 to 5 cm 5 to 6 cm
0.00
0 20 40 60 80 100
Distance from the center of the gap (cm)
Figure 4.3.3-12 Nuclear heat distribution in the direction parallel to the VV surface with
5 manifolds (case 2)
0.70
0.60
0.50
0.40
0.30
0.20
0.10
0.00
0 to 1 cm 1 to 2 cm
2 to 3 cm 3 to 4 cm
4 to 5 cm 5 to 6 cm
0 20 40 60 80 100
Distance from the center of the gap (cm)
Figure 4.3.3-13 Nuclear heat distribution in the direction parallel to the VV surface with
2 manifolds (case 3)
Distribution of the nuclear heat integrated over the front shell thickness (6-cm) is shown in
Figure 4.3.3-14 for the cases of 1(V-gap),2 (5 manifolds)and 3(2 manifolds).
Nuclear Analysis Report Page 75

ITER G 73 DDD 2 01-06-06 W0.1
3.5
3.0
2.5
2.0
1.5
1.0
0.5
0.0
2-manifolds
V-gap
5-manifolds
0 20 40 60 80 100
Distance from the center of the gap
(cm)
Figure 4.3.3-14 Nuclear heating distribution integrated over the vacuum vessel
thickness
4.3.4 Heterogeneity Effects of the Vacuum Vessel on the Inboard TF coil Leg
Nuclear Heating 1
In radiation shielding analyses performed with ITER 3-D basic model 2 , the vacuum vessel is
modelled with three-layer structure; namely with inner and outer shells and “filler region”
between them. The “filler region” has detail structure consisting of radial ribs, borated
stainless steel blocks and cooling water. The preliminary analysis conducted for the outboard
vacuum vessel implied that the difference on the TF coil nuclear heating between the cases of
the simple three layer modelling and of heterogeneity in the filler region is significant.
According to the drawings prepared by designers, heterogeneous model on the inboard
vacuum vessel has been made and investigation of effect of the detail structure on nuclear
heating in the TF coil inboard legs, is conducted. The nuclear heating in the TF coil inboard
legs dominates (8.4 kW with ITER 3-D basic model) in the total ( around 13 kW) which is
close to the design limit (13.7 kW in DRG1). This section describes the effect of the
heterogeneity and shows possible solutions to prevent significant increase of the nuclear
1 H. Iida , L. Petrizzi , V. Khripunov, F. Wasastjerna and G. Ruvutuso, “Shielding Efficiency Problem of the
Present Vacuum Vessel Design and Possible Solutions”, NAG-173, January 2001
2 1. G. Ruvutuso, H. Iida and L. Petrizzi, NAG-168-14-11-00, "Updated of Basic 3-D model of ITER for Monte
Carlo nuclear analyses with MCNP".
Nuclear Analysis Report Page 76

ITER G 73 DDD 2 01-06-06 W0.1
heating in TF coil inboard legs. In this analysis the MCNP has been used and statistic of
results are very good (fsd:1-3%) for all cases.
4.3.4.1 Calculation Models
Vacuum Vessel
Y
Z
Filler Region
Plasma
X
Figure 4.3.4-1 A partial model with heterogeneous structure in the inboard vacuum
vessel
The model shown in Figure 4.3.4-1 is a partial 3-D model used in this study. It is locally
detailed model and has heterogeneous structure in the filler region of the vacuum vessel. The
model has reflection boundaries on its top and bottom (in Z direction) and on the both sides
(in Y direction). On the position of outboard blanket, a reflector (80 % SS and 20 % H2O) is
placed.
A zoomed-up picture of the model is shown in Figure 4.3.4-2. The model has one inboard
blanket module with 2 cm gaps on its edges (toroidal and poloidal edges). Three ribs of 4-cm
thickness are located in the filler region. A water layer of 3.7-cm width is provided just
behind the inner shell. The rest of the space in the filler region is occupied by borated steel
blocks (4 cm x ~ 17 cm) and water. The blocks are arranged in four layers and supported by
the ribs. Figure 4.3.4-3 shows a horizontal cross section of the model and Figure 4.3.4-4 that
of the simplified three-layer model correspond to the ITER 3-D basic model.
Nuclear Analysis Report Page 77

ITER G 73 DDD 2 01-06-06 W0.1
Thermal
Shield
Figure 4.3.4-2 Detail of the Vacuum Vessel
Blanket
TF Coil
Inboard Leg
Vacuum
Vessel
Figure 4.3.4-3 Horizontal Cross Section of the 3-D model which simulate proposed
design of the Vacuum Vessel
Nuclear Analysis Report Page 78

ITER G 73 DDD 2 01-06-06 W0.1
Hoomooggeenneeoouuss Mi ixxt tuurree
1) 3-D Basic Model
B.S.S: 60 %
H2O: 40%
2) Proposed Design
Structure
B.S.S: 72.36 %
H2O: 27.64%
Figure 4.3.4-4 Horizontal Cross Section of the 3-D model with a homogenized Layer
between the two Shells of the Vacuum Vessel
(This is the same modeling as used in ITER 3-D basic model)
4.3.4.2 Comparison of the Nuclear Heating Values for the Proposed Design and the
ITER 3-D Basic Model
Nuclear heating values obtained with using the partial 3-D model are compared in Table
4.3.4-1 with different degree of homogenising approximation. The simplest approximation is
three layer model which correspond to the ITER 3-D basic model. The model with well
simulation of heterogenuity gives 60 - 70 % larger nuclear heating in the TF coil inboard legs
than expected with the 3-D basic model. When we homogenise the filler region between the
inner and outer shells, we have different mixture (water content) from that specified in the 3-
D basic model. Table 4.3.4-1 also gives a nuclear heating value for the homogenised
proposed design case, which gives ~ 20% higher value than the 3-D basic case.
Nuclear Analysis Report Page 79

ITER G 73 DDD 2 01-06-06 W0.1
Table 4.3.4-1 Comparison of the Nuclear Heating Values for the Present Design and
3-D Basic Model
Model
The same mixtur and the same
homogenisation with the ITER 3-D Basic
Model (Figure 4.3.4-4)
Heterogeneous model with Proposed Mixture
( Figure 4.3.4-3)
Proposed Mixture homogenised over the
region between shells (Figure 4.3.4-4)
Proposed Mixture homogenised over the
region between shells except radial ribs
Nuclear Heating in TF
Coil inboard Leg
(W)* kW**
Ratio
33.7 8.4 1.0
57.7 14.4 1.71***
40.7 10.1 1.21
44.3 11.0 1.31
*: Heating in the part just behind one blanket module which is ~109 cm x ~ 130cm.
**: Normalized to 8.4 kW in the inboard TF coil legs
***: This value is slightly exaggerated by the fact that the model used in this study employs old
and thicker (5.5 cm) thermal shield than the latest design. When its thickness is reduced to 2.5
cm the value of this ration become 1.62.
The higher nuclear heating in the case of the proposed vacuum vessel design, comes mainly
from the fact that the amount of cooling water in the filler region is reduced from that
specified (60% SS and 40% H2O) 1 in the Basic 3-D model. It enhanced significantly
heterogeneity effect of the vacuum vessel structure on the nuclear heating in the TF coil.
It is well known that the some amount of light isotope, such as hydrogen or carbon, is
necessary in a mixture to provide high shielding efficiency for fast neutron. Significant
amount of parametric analyses for optimising mixture composition has been conducted in the
past. Figure 4.3.4-5 shows an example of such analyses 2 . Typically, water volume fraction
from 15 % to 40 % (then 85% to 60% of Steel) in the mixture of stainless steel and water
gives “effective shielding region” and outside of this range is a “dangerous region”. In the
Table 4.3.4-2, the values of water volume fractions are summarised for the mixtures used in
this study averaged in the whole thickness of the vacuum vessel and/or “filler region”
between two shells of the vacuum vessel.
As we can see in Table 4.3.4-2, the specified mixture in the 3-D model gives almost
optimised point (75 % SS & 25 % Water) when it is averaged over the whole vessel
thickness. On the other hand, the water fraction of the proposed design is almost on the edge
of “effective shielding region” for the case of homogenised mixture.
In the proposed design of vacuum vessel structure, the large part of water in the filler region,
localise near the inner shell leaving other part waterless. This caused a sharp increase of
nuclear heating in the TF coil inboard legs. Then, increase of water fraction can be a possible
solution to improve present design of the vacuum vessel and to avoid sharp increase of TF
1 1. R.T. Santoro, V. Khripunov, H. Iida, , NAG-101-98-06-17-CDR "ITER NUCLEAR ANALYSIS REPORT"
June 1998
2 1. V. Khripunov, R. T. Santoro, H. Iida, NAG-11-11-13-96, "Discrete Ordinates Analysis of the Nuclear
Performance of the ITER Equatorial Ports"
Nuclear Analysis Report Page 80

ITER G 73 DDD 2 01-06-06 W0.1
coil nuclear heating from expected value with 3-D Basic model. This can be seen
quantitatively in the following section.
100000
10000
1000
100
Heat in 20 TFC
10
0.00 0.25 0.50 0.75 1.00
SS-Volume Fraction in the VV
SS-316
B(1%weight) * SS
B(2%weight) * SS
Figure 4.3.4-5 Nuclear Energy Release in 20 TFC as a Function of Steel Fraction in 65cm
Vacuum Vessel 6 ; (Absolute values of increasing factor are not applicable for the present
case because of different conditions such as shield thickness)
Average in
the Block
Region
Average
over the
Filler
Region
Average
over the
whole VV
thickness
Table 4.3.4-2: Composition of mixtures of steel and water for
the vacuum vessel in this study
3-D Basic
Model
Mixture
Proposed
Design
Reduced Metal Cases
75 % 62.5 % 50 %
Fraction of
Borated
0.859 0.644 0.537 0.430
Stainless Steel
Fraction of
Water
Total Number
Density
-
-
-
0.141 0.356 0.463 0.571
9.3086E-2 9.4909E-2 9.5821E-2 9.6733E-2
Fraction of
Borated
Stainless Steel
Fraction of
Water
Total Number
Density
Fraction of
Borated
Stainless Steel
Fraction of
Water
0.6 0.739 0.580 0.5 0.420
0.4 0.261 0.420 0.5 0.580
9.5284E-2 9.4103E-2 - - -
0.742 0.832 0.729 0.677 0.626
0.258 0.168 0.271 0.323 0.374
Note: The volume fraction of the ribs in the “filler region” is 10.1 %. This part is SS316-IG in
heterogeneous model and borated steel in the homogenised model.
The volume fraction of the 3.7-cm water layer in the “filler region” is 15.7 %
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ITER G 73 DDD 2 01-06-06 W0.1
4.3.4.3 Improvement of shielding efficiency by increasing water fraction (or
decreasing steel fraction)
Under the constraints of geometry given by the designers, which require three radial ribs per
a sector and 3.7 cm thick water layer just behind inner shell of the vacuum vessel, parametric
analyses have been conducted reducing stainless steel and increase water volume fraction.
The following approach was considered in reducing stainless steel volume fraction.
1) 1)Remove layers of borated steel blocks one by one from the present design with four
layers. (see Figure 4.3.4-6a and 4.3.4-6b)
2) reduce thickness of borated steel uniformly (see Figure 4.3.4-7)
3) reduce thickness of borated steel with grading (see Figure 4.3.4-8)
The results of the analysis are shown in Figure 4.3.4-9. Increasing volume fraction of water
and then reducing that of steel is very effective to improve shielding efficiency of the vacuum
vessel and optimum point exist in the rage of 25 % -50 % reduction of borated steel in the
filler region from the proposed design. The improving factor does not depend much on the
way of reducing steel volume fraction as far as it makes water distribution closer to uniform.
Only slight difference can be seen among the above approaches and best improvement is
obtained with reducing steel block thickness with grading (approach 3). The improvement
yields 12 % - 18% larger nuclear heating than that for ITER 3-D basic model. It may not be
easy to have much better improvement under the constraints given by the designers.
Figure 4.3.4-6a Model in which the
last layer of steel block is replaced
with water
Figure 4.3.4-6b Model in which the
last and second layers of steel block
are replaced with water
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 4.3.4-7 Model in which the
thickness of steel block is reduced
from 4 cm to 3 cm
70
60
50
40
30
20
10
0
reduce number of Blocks
reduce Block Thickness
Figure 4.3.4-8 Model in which the
thickness of steel block is reduced
with grading
Nonuniform distribution of Block Thickness
3-D Basic Model
Proposed Design (Hetero Model)
Homogenized Proposed Design.
Bad position (remove first layer)
0 0.2 0.4 0.6 0.8 1
Relative amount of Borated SS
Figure 4.3.4-9 The effect of reducing borated steel volume fraction on the nuclear
heating in the TF coil inboard legs
(note: “Bad position” removes the first layer blocks not providing any additional water to the part of filler
region closer to the rear shell)
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ITER G 73 DDD 2 01-06-06 W0.1
Conclusion
Proposed design can be improved by increasing water volume fraction in the filler region
avoiding sharp increase (~ 60 – 70 %) of nuclear heating in the TF coil inboard legs.
However, due to the structural and thermo-hydraulic design constraints that three ribs per
sector and 3.7 cm thick water layer just behind the inner shell are necessary in the region
between two shells of the vacuum vessel, the estimate of nuclear heating in the TF coil
inboard legs should be increased by 10 – 20 % ( 0.8 kW – 1.7 kW) from the value obtained
with ITER 3-D basic model.
4.4 Divertor Cassette
4.4.1 Modelling
The divertor cassette has a complex geometry and is made of two kind of components that
have different purposes: the plasma-facing components for very high heat load removal,
called High Heat Flux Components (HHFCs), and an underlying robust “cassette body”.
A detailed model of the cassette and the HHFCs has been done. Very few approximations
have been introduced: the material dilution has been used only in cases when the achieved
model simplification gives negligible neutron flux variation. The detailed model guarantees
that the poloidal and radial neutron flux variations can be properly calculated. The alternation
of the different material has been reproduced in the model, avoiding any homogenization.
This is very important especially in the front regions of the HHFCs where thin layers of
different materials are present. Some simplifications have been done, using equivalent
thickness layers. For example flat layers of copper and water of equivalent thickness
represent the water tubes of the HHFCs. Full account has been taken of the pumping slots and
the gap in between the cassette, which affect the streaming through the lower ports. Fine cell
subdivision has been done in order to have a detailed poloidal and radial distribution of the
nuclear heating. In the 20˚ model two whole cassettes and two half cassettes are described
(Figure 4.4.1-1, 4.4.1-2) to keep the exact symmetry of the system with respect to the port
divertor.
The cassette itself has been modeled with its two steel layers. The rear steel layer is 70 mm
thick while the front layers are 50 mm thick. The water layers in between is 130 mm thick,
for a total thickness of 250 mm. In some places the overall thickness can be more than that,
where the design foresees more stiffness of the component, like in the positions close to the
attachments to the VV. Steel walls 70 mm thick close the cassette sides, internal ribs are 35
mm thick. There are three cassettes per 20˚ sector, with 10 mm gap in between each of them.
The pumping duct is at the side of outboard side edge of the cassette, with toroidal width 50
mm and 500 mm height. It is a penetration all along the cassette to allow the pumping of the
exhaust plasma.
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ITER G 73 DDD 2 01-06-06 W0.1
Fig. 4.4.1-1 Isomeric view of the ensemble of the cassettes (2 whole 2 halves)
in the 20˚ sector model used for MCNP calculations.
Fig. 4.4.1-2 Isomeric view of the ensemble of the (2 whole 2 halves) in the 20˚
sector model, seen from the top. In yellow, the tungsten coating covering most
part of the HHFCs
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ITER G 73 DDD 2 01-06-06 W0.1
4.4.2 Nuclear heat distribution, damage, Helium production and shielding
capability
The objective of these neutronics analyses is to assess the shielding capability of the cassette
respect to the Toroidal Field Coil, taking into account the streaming through the openings and
the gaps. Moreover a map of the nuclear heating is required for the thermal analyses and
thermal hydraulics sizing. In some locations special response functions have been calculated,
such as the helium production in the manifolds, for their reweldability.
Two pictures of the model are in Figures 4.3.2-1 and 2, in which the segmentation of the
space is shown. Many cells were required to proper model the complicate pattern of the
divertor components. At the same time a fine spatial distribution of the nuclear power
generated is preferable for the thermal analyses. A finer distribution has been calculated by
means of surface segmentation in the tally card. The values of the nuclear power deposited
has been collected in electronic format for the cells of the divertor system, but to understand
the trend a coarser mesh is preferable, as in Figures 4.3.2-3 and 4.
Water manifolds
Figure 4.4.2-1 Poloidal section of the model used for MCNP calculation.
In the HHFCs the power density is strongly varying according to the different orientation of
the cells to the neutron source. The maximum power density on the upper tungsten coating
ranges between 4 ÷ 12 W/cm 3 . The graphite front layer covering the two vertical targets has
much lower power density (0.4 ÷ 0.8 W/cm 3 ), both because graphite is a light element with
low Z, and because the vertical target position is less exposed to the direct neutron source.
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ITER G 73 DDD 2 01-06-06 W0.1
The power density on the rear steel layer which gives stiffness to the HHFCs themselves is
0.5 ÷ 1.3 W/cm 3 .
Figure 4.4.2-2 Poloidal section of the model used for MCNP calculation, through a
plane passing through a cassette sidewall. In this section the pumping slot is visible as
well the reinforcement to sustain the cassette to the outer rail.
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ITER G 73 DDD 2 01-06-06 W0.1
0.6
0.9
8
1.8
0.4
9
5
0.6
1.8
12
2.7
1.6
Figure 4.4.2-3 Poloidal contour of the nuclear heating in different materials of the
HHFCs. Values are W/cm 3 . In red the tungsten layer. In blue the Carbon, in black the
steel.
In the cassette itself the poloidal variation of the power density is smoother, due also to its
more regular shape compared to the HHFCs. The biggest difference is between the inboard
recess zones and the outboard zones. In the front steel layer the power density is between
0.34 and 0.8 W/cm 3 , the rear layer 0.01 ÷ 0.2 W/cm 3 .
Nuclear Analysis Report Page 88
7
1.2
4
0.8
0.8
0.6
8
1.8

ITER G 73 DDD 2 01-06-06 W0.1
0.04
0.20
0.1
0.34
0.8
0.6
0.09
0.78
Figure 4.4.2-4 Poloidal contour of the nuclear heating in the three different steel
layers composing the cassette. Values are W/cm 3 . In red the rear layer. In black the
front one.
The Helium production has been calculated in some places to assess the reweldability, in the
manifolds and the attachments. The obtained values, shown in Figure 4.3.2-5, are given for a
continuous irradiation time of 1 year at 500 MW. The actual fluence of ITER is half of that
approximately. Helium production ranges between 0.7 and 15 appm. The value depends
mostly on the high energy flux, then on the direct sight of the neutron source (the plasma).
The reweldability limit is about 3 appm for thin plate or tube welding 1 and 1 appm for thick
plate or tube. Present design incorporates the cassette replacing scheme so that the rewelding
parts should have always less helium production than the above limits.
The maximum dpa on Carbon tiles of the vertical target is 1.2 for 1-year full irradiation time.
1 Design Requirements & Guidelines Level 1
0.6
Nuclear Analysis Report Page 89
0.05
0.4
0.2
0.8
0.05
0.03
0.04
0.6
0.01

ITER G 73 DDD 2 01-06-06 W0.1
5
2
11
13
10
7
14
Figure 4.4.2-5 Poloidal contour of the helium production in steel in some locations
(manifolds and attachments). Values are appm for 1 year full time irradiation.
Integral values of nuclear power are important as well. Table 4.4.2-1 gives a summary of the
integral nuclear heating values
Table 4.4.2-1 Integral nuclear power on the 54 cassettes [MW]
Outer target 21.7
Inner Target 9.5
Dome and liner 17.3
Divertor’s cassette 9.8
TOTAL POWER 58.3
4
The cassette body should give enough shielding respect to the Toroidal Field Coils (TFC).
The requirements are that the power density on the steel case should be < 2 mW/cm 3 , 1
Nuclear Analysis Report Page 90
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15
5
13

ITER G 73 DDD 2 01-06-06 W0.1
mW/cm 3 on the winding pack of the superconductor. The total overall nuclear heating should
be < 13.7 kW for all the 18 TFCs 1 .
A shielding analysis has been conducted, to be sure that the cassette gives enough shielding
in its reference design. The nuclear heating has been calculated in a limited poloidal
extension of the TFC: only the part behind the divertor has been described. The nuclear
power deposited in the TFC is 380 W for the 18 coils. In the model the ports located in the
divertor regions are closed, no streaming is considered through those openings. In that case
then the contribution coming from the transport in the cassette is not mixed and overwhelmed
by the contribution coming from the streaming through the ports.
The power density on the steel case and in the conductor is well below the limit (one order of
magnitude less).
The values are consistent with what has been independently calculated in the frame of the
general analysis of ITER machine 2 .
An important topic is the analysis of the streaming through the penetrations of the divertor
components. This has a strong impact on the radiation level in the divertor port, which has
been fully analysed in the report 3 . It has been shown that there are two major sources of
streaming of radiation in the divertor port opening which are responsible for the relative high
activation in that area are. They are the skew gap between the divertor and the bottom part of
the outer blanket and the pumping slot in between the cassettes.
1 Design Requirements & Guidelines Level 1
2 G. Ruvutuso and H. Iida, NAG-159-08-06-00, “Three-Dimensional model of the ITER-FEAT reactor for
Monte Carlo nuclear analyses with MCNP”.
3 G. Ruvutuso, L. Petrizzi and H. Iida, NAG-164-07-08-00, “Radiation Conditions inside the divertor ports”.
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ITER G 73 DDD 2 01-06-06 W0.1
5 Port Analysis
In this chapter results of the 3-D analyses for the ports are summarised, except for those used
for diagnostic purpose. Analyses for diagnostics ports are reported in Chapter 6. Main
interests of the analyses are dose rate after shutdown and nuclear responses in the magnet
system around ports. Shutdown doses presented in this chapter were obtained with the one
step method (see Appendix B) except LH and Test module ports analysis ( sections 5.5 and
5.6).
5.1 ECH Upper port 1
The model of the upper port ECH launcher was based on drawings obtained in January to
March 2001. This was inserted in the ITER 3D basic model. Like for the equatorial port ECH
launcher, the lack of symmetry made it necessary to model both halves, connected by a
periodic boundary condition.
Figure 5.1-1 shows an elevation view of the port. Figure 5.1-2 shows a plan view of the right
half.
There is additional shielding below the front part of the launcher, not visible in Figure 5.1- 1
because it does not lie in the picture plane. Even so, the rather open area here is a weak spot
in the shielding.
Figure 5.1-1 Elevation view of upper port with ECH launcher
1 F. Wasastjerna, NAG-174, “3-D Analyses for the ICH, ECH and LH Ports in ITER-FEAT”, May , 2001
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.1-2 Plan view of right half of upper port with ECH launcher
5.1.1 Nuclear heating in the magnet system
Nuclear heating in the magnet system around the port is given in Table 5.1-1 and 5.1-2 and is
smaller than those reported in section 4.5 (Table 4.5.1-1), which is obtained with the ITER
3D basic model. The basic model has a large void at the root of the upper port but in the
analysis of ECH port, the void is filled with the pieces of shield block. Even though the
mechanical design of that part is still under way, the shield configuration employed in this
analysis is more realistic than the basic model. The calculated heating value is 24.7 W/port. If
we apply for all 18 upper ports , the total heating for the upper ports is 445 W which is much
smaller than that obtained with ITER basic model (1.4 kW, see section 4.5)
Table 5.1-1 Nuclear heating in the TF coil around the Upper ECH port
Nuclear
Heat
(W)
Coil case Insulator Winding
Pack
Sub total Intercoil
structure
13.4 0.26 6.95 20.6 4.1
Total
24.7/port
74.1 (3port)
Table 5.1-2 Nuclear heating in the PF coil around the Upper ECH port
Nuclear
Heat
(W)
1) Shutdown dose rates
PF#1 PF#2 PF#3 Total
0.13 0.71 0.63
1.47/port
The shutdown dose rates are shown in Figures 5.1-3 through 5.1-4. Due to insufficient time
for calculations the statistics are in many cases not necessarily good. For the large cells
around the rear part of the port, outside the port walls, the fsd (fractional standard deviation)
has relatively reasonable values, mostly below 0.2. In the smaller cells farther forward on the
outsides of the port walls, the fsd values are 0.2 – 0.4. In the figures the values which have
smaller fsd than 0.2 are presented.
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ITER G 73 DDD 2 01-06-06 W0.1
Only dose rates above the port seem to be lower or close to the design limit. Dose rates at
both sides and below the port are significantly higher than the limit of 100 μSv/h. Especially,
those around the left port are high suggesting that the most problematic place exists in the
lower part of the left port.
Similar results are also obtained by the RF HT 1 . Significant improvement of the shield
configuration of the port including ECH launcher is necessary.
101
120
63
278
174
48
103
Figure 5.1-3 Dose rates in μSv/h 10 6 s after shutdown behind, above and below port
1 Neutronic Analysis of the ITER Vacuum Vessel/Cryostat Environment. Report of RF HT for the 4 th quarters
2000. JF-04-00/4. December 2000.
Nuclear Analysis Report Page 94
37
50
31

ITER G 73 DDD 2 01-06-06 W0.1
255
Figure 5.1-4 Dose rates in μSv/h 10 6 s after shutdown right of port
822
660
Figure 5.1-5 Dose rates in μSv/h 10 6 s after shutdown left of port
5.2 NBI Port 1
The #4 and #5 equatorial ports are used for the NBI for plasma heating. Diagnostic neutral
beam injector is also installed in the #4 port as well. Since the neutral beam lines in these
ports are completely void, the NBI ports can provide a major concern of neutron streaming
from tokamak machine and careful shield design is required. When sufficient thickness of ~
60 cm is allowed for the NBI port wall, there is no problem for obtaining low enough dose
rate (< 100 micro Sv/h) as well as nuclear heating in the cryo-temperature components (~
50W/port) around the port. However, NBI port will have interference with TF coils and/or PF
coil support allowing only limited space for the port wall. The thinnest part of the wall has a
thickness of ~45 cm. Thus, detailed and exhaustive analysis is required.
5.2.1 Calculation model
102
Figure 5.2-1 shows 3-D view of the Model created by SABRINA code 1 with the geometry
input data for MCNP. The upper half is made transparent for better view of inner part.
1 S. Sato, Y. Ohara and M. Akiba, “Radiation streaming analysis through equatorial port”, JP HT report of
D469-JA, 2001, June
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56
219 67

ITER G 73 DDD 2 01-06-06 W0.1
Figures 5.2-2 and 5.2-3 show a horizontal and vertical cross sections of the model.
Especially, Figures 5.2-3(a),(b),(c) show thickness of shielding pieces which locate in a
critical position.
Bio-shield
D-NBI
NBI-2
Plug
NBI-1
PF Coils
Inboard Blanket
Divertor
Vacuum Vessel
TF Coil
Center Solenoid
Figure 5.2-1 NBI 3-D MCNP model by SABRINA code (upper half is cut for making
inner part visible)
1 Kenneth A. Van Riper, SABRINA User’s Guide, White Rock Science, P.O. Box 4729, White Rock, NM,
87544 USA.
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ITER G 73 DDD 2 01-06-06 W0.1
Outboard blanket
Inboard blanket
PF coil support structure
Vacuum vessel
TF coil
Plasma
NBI-2 port
NBI-1 port
Cryostat
Figure 5.2-2 Horizontal cross section of the NBI 3D model
TF coil inter-coil structure
Insert part of
D-NBI port
NBI-1 port
Vacuum vessel
38.9 cm
PF coil(#3)
28 cm
Biological shield
28.2 cm
32 cm
15 cm
10 cm
Cryostat
D-NBI port
Biological
shield
Figure 5.2-3 (a) Vertical cross section of the NBI 3D model (NBI-1;#4 port)
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ITER G 73 DDD 2 01-06-06 W0.1
TF coil inter-coil structure
PF coil
27 cm
20 cm
D-NBI port
Insert part of
NBI-1 port
38 cm
Biological shield
10 cm
15 cm
Vacuum vessel
Cryostat
Figure 5.2-3 (b) Vertical cross section of the NBI 3D model (DNB;#4 port)
Blanket
NBI-1 port
Vacuum vessel
TF coil inter-coil structure
PF coil
Biological shield
28.2 cm
32 cm
15 cm
10 cm
38.9 cm
28 cm
Cryostat
Figure 5.2-3 (C) Vertical cross section of the NBI 3D model (NBI-2;#5 port)
5.2.2 Results of the analysis
5.2.2.1 Nuclear heating and insulator dose rate in the magnet system
Nuclear heating in the magnet system around the port is given in Table 5.2-1 is small enough
giving no significant problem on the magnet system. Table 5.2-2 shows insulator dose in the
TF coil assuming 0.5 MWa/m 2 of fluence averaged over entire first wall surface. The
insulator specific weight of 1.699 g/cm 3 is assumed. The absorbed energy by the insulator is
much less than the limiting value of 10 E+6 Gy.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 5.2-1 Nuclear heating in the TF coil around the NBI port
Nuclear Heat
(W)
Coil case Insulator Winding Sub total Intercoil Total
Pack
structure
31.2 3.34E-01 6.14 37.7 41.1 79
160 (2port)
Table 5.2-2 Insulator dose in the TF coil around the NBI port
total
(Gy)
Insulator 4.50 E+03
5.2.2.2 Neutron fluxes and dose rate after shutdown around the port
Dose rates at 10E+6 sec ( ~ 11 days) after shutdown on the equatorial plane between the two
NBI ports are shown in Figure 5.2-4. The dose rates just outside the wall of NBI-1 is very
high giving the maximum around 3400 micro Sv/h. However, PF coil support is providing
shielding function also while is preventing to allow thicker wall for the port. The dose rate
near the cryostat is reduced by a factor of ~ 6. It is still significantly higher than the target
dose of 100 micro Sv/h for hands-on maintenance. According to the maintenance
requirement, the equatorial plane is not the place to access for maintenance of components,
such as break boxes of magnet system. The space along the cryostat between the port and
PF#3 coil (or PF#4 coil) is a place relevant to the maintenance works and required for access.
Figures 5.2-5 to -10 show the dose rates at these places. They are generally lower than those
at equatorial plane but still higher than the target. Especially, at the space just below the PF#3
coil and just above the PF#4 coil, significantly high dose spots exist (300–900 micro Sv/h).
These dose rates are mainly contributed by the radiation leakage through the thin part of port
wall which locates near the cryostat (for example; see Figure 5.2-6 #1 cell). Some
improvement of shielding structure on this part is necessary to reduce the dose rate below 100
micro Sv/h.
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ITER G 73 DDD 2 01-06-06 W0.1
unit: μSv/hour
(fsd: %)
270(13.8)
261(8.82)
505(9.34)
561(12.3)
470(7.02)
608(5.80)
552(11.5)
426(7.39)
510(10.9)
NBI-2 port
Cryostat Biological shield
311(8.20)
347(7.29)
220(7.84)
348(12.9)
555(10.5)
302(8.56)
NBI-1 port
542(9.84)
TF coil 568(7.71)
527(14.7)
3360(6.57)
PF coil support structure
Figure 5.2-4 Shutdown dose rate distribution (10 6 sec ) on the equatorial plane between
the two NBI ports
unit: μSv/hour
64.1(22.3)
(fsd: %)
188(17.6)
102(20.5)
PF coil(#3)
175(17.4)
220(19.1)
PF coil support structure
TF coil
TF coil inter-coil structure
Vacuum vessel
Blanket
89.3(24.0)
167(20.1)
155(17.9)
111(15.3)
107(15.6)
80.7(22.9)
103(20.8)
109(18.7)
Cryostat
89.3(13.3)
197(15.7)
Biological shiled
145(18.0)
147(13.8)
84.2(17.4)
146(17.3)
173(17.6)
172(16.8)
108(18.0)
Figure 5.2-5 Shutdown dose rate (10 6 sec ) on the level of PF#3 coil above the NBI ports
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ITER G 73 DDD 2 01-06-06 W0.1
66.7(16.8)
PF coil(#3)
68.5(17.0)
NBI-1 port wall
449(8.48)
102(13.8)
172(16.8)
173(17.6)
175(21.0)
155(13.5)
131(11.5)
#1 Cell
96.2(23.6)
unit: μSv/hour
(fsd: %)
Biological shield
Cryostat
267(11.4)
108(18.0)
Figure 5.2-6 Shutdown dose rate (10 6 sec ) around the NBI1 platform for PF#3 coil
maintenance
60.4(23.3)
PF coil(#3)
542(10.9)
220(19.1) 129(19.7)
253(22.7)
175(17.4)
109(23.7)
80.7(22.9)
115(23.4)
Biological shield
Cryostat
unit: μSv/hour
(fsd: %)
NBI-2 port wall 897(10.6)
193(16.1)
260(12.8)
Figure 5.2-7 Shutdown dose rate (10 6 sec ) around the NBI2 platform for PF#3 coil
maintenance
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ITER G 73 DDD 2 01-06-06 W0.1
PF coil(#4)
PF coil support structure
TF coil support structure
TF coil
TF coil inter-coil structure
Vacuum vessel
Blanket
123(16.2)
44.2(20.5)
175(24.5)
237(14.1)
135(14.1)
122(17.3)
137(13.1)
143(17.2)
Cryostat
174(17.6)
115(12.6)
94.5(27.3)
198(15.3)
165(16.3)
191(19.6)
108(15.7)
195(14.1)
183(16.9)
163(15.5)
165(22.1)
81.0(22.5)
113(17.0)
unit: μSv/hour
(fsd: %)
Biological shiled
103(18.6)
Figure 5.2-8 Shutdown dose rate (10 6 sec ) on the level of #4 PF coil below the NBI ports
NBI-1 port wall
395(16.0)
PF coil(#4)
431(22.9)
194(15.7)
243(20.7)
138(12.8)
unit: μSv/hour
(fsd: %)
Cryostat
103(18.7)
Biological shield
39.6(20.4)
187(25.9)
81.0(22.5)
113(17.0)
63.9(20.7) 198(24.0)
Figure 5.2-9 Shutdown dose rate (10 6 sec ) around the NBI1 platform for PF#4 coil
maintenance
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ITER G 73 DDD 2 01-06-06 W0.1
NBI-2 port wall
321(19.3)
PF coil(#4)
90.2(15.6)
299(17.3)
135(14.1)
166(19.2)
220(22.9)
143(17.2)
115(11.2)
155(18.4)
31.7(31.7)
127(25.3)
Biological shield
94.5(27.3)
Cryostat
unit: μSv/hour
(fsd: %)
Figure 5.2-10 Shutdown dose rate (10 6 sec ) around the NBI2 platform for PF#4 coil
maintenance
5.3 ECH Ports 1
The model of the equatorial ECH launcher was based on drawings obtained in
September/October 2000. This was inserted in the ITER 3D basic model, which included the
inter-coil structure but not the blanket cooling manifolds. A simple model of the cryostat and
bioshield was added.
24 circular waveguides with an inner diameter of 63.5 mm were modelled in the launcher.
Each waveguide had 3 mitre bends. Due to the structure of the general models, with two half
equatorial ports, the launcher had to be split into two parts connected by a periodic boundary
condition. The total thickness of the box enclosing the launcher was 130 mm, consisting of
40 mm steel + 50 mm water + 40 mm steel. The gap between this box and the port walls was
20 mm thick, with an offset of 30 mm at the dogleg, except for the rear end, near the flange
attaching the launcher to the port. Here the gap was reduced to 5 mm over a length of 150
mm.
Figure 5.3-1 shows elevation and plan views of the launcher and its immediate
neighbourhood. The dose rates for the nominal design with a 5 mm gap thickness over the
last 150 mm, are shown in Figures 5.3-2 to 5.3-6. Fractional standard deviations were below
0.2 for all the dose rates shown.
1 F. Wasastjerna, NAG-174, "3-D Analyses for the ICH, ECH and LH Ports in ITER-FEAT",May ,2001
Nuclear Analysis Report Page 103

ITER G 73 DDD 2 01-06-06 W0.1
The nuclear heating of the cryogenic components from one port was 2.73 W for the TFCs,
0.9 W (with rather bad statistics) for the PFCs and 1.22 W for the intercoil structures, totaling
4.85 W for one port or 87.4 W for 18 ports.
Figure 5.3-1 Elevation and plan views of two half equatorial ports with ECH launchers
123 43 26 23
Figure 5.3-2 Dose rates in μSv/h 10 6 s after shutdown behind the equatorial ECH port
for the nominal design
Nuclear Analysis Report Page 104
12

ITER G 73 DDD 2 01-06-06 W0.1
The shield plug of present design is too heavy for transfer cask for maintenance. It should be
less than 40t. There is a need of parametric study to confirm that lighter alternative designs
can be possible without deteriorating shielding performance. In addition, with this design, the
dose rate in the cavity in the rear of the launcher is slightly higher than the limit of 100 μSv/h
requiring a small change in the design.
An important fact was that essentially no neutrons came through the plug itself. In fact, at the
rear end of the plug the direction of the net neutron current was into the plug. Thus it seemed
feasible to replace much of the steel in the plug with water, making it lighter though less
effective as a shield. Some calculations were performed to study this, dividing the plug into 3
sections and varying the composition of the 2 rear sections. The boundaries between the
sections were located 871 and 2050 mm from the front surface of the flange.
Table 5.3-1 shows how the shutdown dose rate in the cavity in the rear of the launcher
changes when the design is changed in various ways. In most cases (except case 5 of thinner
port wall) the dose rate elsewhere doesn’t change much.
Cases 6 and 7 are the those of reducing stainless steel in the plug in order to reduce plug
weight. They show that varying the composition of the middle and rear parts of the plug does
not have much effect on the dose rate, at least so long as the changes do not exceed what was
modelled here.
A comparison of cases 1 and 2 shows that, as expected, thickening the rear end of the frame
(the part surrounding the cavity) decreases the dose rate by 1/2 ~ 1/3. In place of changing
the thickness of the frame, changing the composition of the end part of the frame is
conceivable (case 1,3,4 and 5). From those cases, replacing water in the end part of the frame
seems to give slightly lower dose rate in the cavity.
In the case with the port walls reduced to 100 mm of steel (case 5), the dose rate over most of
the surface of the port walls exceeds 200 μSv/h, whereas for the 200 mm steel-water-steel
wall it’s mostly below 100 μSv/h. The dose rate at the outer surfaces of the port walls is
shown in Figures 5.3-3 through 5.3-6. The values in parentheses apply to case 5, the others
are for case 7, but the values for cases 1 through 4 and 6 are not drastically different from
those for case 7. For these dose rates also the fsd is mostly below 0.2.
Table 5.3-1 Effects on the shutdown dose rate in the cavity of varying design
parameters
Plug composition Frame Dose rate
Case %SS in: Thicknesses (mm) in fsd Comments
front-middle-rear Total SS-H2O-SS rear cavity
1 80-80-80 130 40-50-40 123 0.08 Nominal design
2 80-80-80 200 60-80-60 41 0.10
3 80-80-80 130 40-30-60 93 0.08
4 80-80-80 130 130-0-0 85 0.15
5 80-80-80 130 130-0-0 108 0.12 10 cm port walls
6 80-50-0 130 130-0-0 91 0.15 Lighter plug
7 80-30-30 130 130-0-0 87 0.13
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ITER G 73 DDD 2 01-06-06 W0.1
221(795) 116(300) 58(204) 62(219) 60(170)
Figure 5.3-3 Dose rates in μSv/h 10 6 s after shutdown along the upper surface of the
port (values for 100 mm single-layer port walls in parentheses)
357(805)
Figure 5.3-4 Dose rates in μSv/h 10 6 s after shutdown along the right side of the port
(values for 100 mm single-layer port walls in parentheses)
187(503) 113(326) 47(149) 38(148) 43(154) 27(68)
Figure 5.3-5 Dose rates in μSv/h 10 6 s after shutdown along the lower surface of the port
(values for 100 mm single-layer port walls in parentheses)
716(1380)
201(390)
196(543)
82(243)
102(362)
71(210)
79(255)
53(213)
78(284)
50(77)
38(105)
49(85)
Figure 5.3-6 Dose rates in μSv/h 10 6 s after shutdown along the left side of the port
(values for 100 mm single-layer port walls in parentheses)
Nuclear Analysis Report Page 106

ITER G 73 DDD 2 01-06-06 W0.1
5.4 ICRF Port 1
The ICH is installed in equatorial ports #13 and #15. In the calculation model, an ICH
launcher based on the drawings as April 2000 [drawing number 51 0067 0001], was inserted
in the equatorial half-port in Sector 3 of the ITER basic model (see Figures 5.4-1 and 5.4-2).
The equatorial half-port in Sector 1 contains a simple shielding plug, but the details of this
are irrelevant, since particles were killed in this port. Likewise, particles in many other parts
of the system were killed, in such a way that the calculated results give the contribution only
from neutrons emerging through the ICH port. A closure plate 2.2 cm thick is at the end of
the port. The general model in which the ICH launcher was inserted was that available in
March 2000, with, among other things, the inter-coil structure missing. A simple model of the
cryostat and bioshield was added to account for reflection in these components.
The ICH launcher plug provides rather good shielding. The streaming through the gap
between the plug and the port walls is the dominant contributor to the neutron flux
responsible for activation. With the initial model, shutdown dose rates were somewhat too
high, but reducing the gap thickness to 5 mm over the last 150 mm and adding 100 mm to the
length of the plug reduced the shutdown dose rate to the values shown in Figures 5.4-3
through 5.4-6. These are generally below the limit of 100 µSv/h outside the port and
coverplate. The fractional standard deviations mostly lie between 0.1 and 0.2.
The total heating in the cryogenic components was also calculated. The heating in the TF
coils was 0.35 W for half a port and in the PF coils 0.19 W. Adding this and doubling to give
the result for a whole port gives 1.07 W for all cryogenic components except the missing
intercoil structures. For 18 ports the corresponding figure would be 19.3 W. it is likely that
this figure is underestimated because of the absence of the intercoil structures and the fact
that only the most important poloidal part of the TFC was included in the tally, but it is
unlikely that the true result would be more than a factor of 2 higher.
Figure 5.4-1 ICRH 3D MCNP model
by SABRINA
Figure 5.4-2 Vertical cross section
1 F. Wasastjerna, NAG-174, “3-D Analyses for the ICH, ECH and LH Ports in ITER-FEAT”,May ,2001
Nuclear Analysis Report Page 107

ITER G 73 DDD 2 01-06-06 W0.1
62 22 21 30
196 144 76
183
65 36 24 27
Figure 5.4-3 Dose rates in μSv/h 10 6 s after shutdown in the rear of the ICH launcher
and above and below the port
Nuclear Analysis Report Page 108
54

ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.4-4 Dose rates in μSv/h 10 6 s after shutdown at the corner of the port. The
values shown are averages over locations at the top, side and bottom
93
37
22
Figure 5.4-5 Dose rates in μSv/h 10 6 s after shutdown at the sides of the port
Nuclear Analysis Report Page 109
27
88
51
36
29

ITER G 73 DDD 2 01-06-06 W0.1
28 16
Figure 5.4-6 Dose rates in μSv/h 10 6 s after shutdown bewteen the port and the bioshield
(cryostat modeled only crudely)
5.5 LH Equatorial Port 1
The LH launcher was also modelled on the basis of drawings obtained in January to March
2001 and inserted in the ITER 3D basic model. The geometry and the results are shown in
Figures 5.5-1 and 5.5-2.
The shield block at the rear of the launcher is penetrated by waveguides, but they are not
visible in the elevation view since they do not lie in the picture plane.
The analysis is still preliminary for dose rates. However, a reasonably good estimate of the
neutron flux above 1 MeV was obtained, with fractional standard deviations below 0.16 for
all locations where results are shown in the figures, with the exception of the third and fourth
value above the port where the fsd was about 0.32. This flux was converted to a shutdown
dose rate by using a conversion factor of 1.33*10 -5 (μSv/h)/(n/cm 2 s), obtained for locations at
the outer surfaces of the port for the equatorial ECH launcher. The dose rates are italicized as
a reminder of this. It is not likely that the use of a conversion factor leads to really large
errors in this case, since the geometry of the calculation from which the conversion factor
was derived was similar. If one wishes to be conservative, one can apply a safety factor of
1.7, which is the approximate ratio of the maximum conversion factor in any relevant cell in
the equatorial ECH calculation to the average used here.
Even with such a conversion factor, the estimated dose rate would remain below 100 μSv/h
except near the root of the port walls, where hands-on maintenance seems unlikely.
1 F. Wasastjerna, NAG-174, “3-D Analyses for the ICH, ECH and LH Ports in ITER-FEAT”,May ,2001
Nuclear Analysis Report Page 110
12
11

ITER G 73 DDD 2 01-06-06 W0.1
174 71 52 38 33
173 67 33 25 26
Figure 5.5-1 Elevation view of LH port with dose rates in μSv/h 10 6 s after shutdown
156 63 41 27 23
Figure 5.5-2 Plan view of LH port with dose rates in μSv/h 10 6 s after shutdown
Nuclear Analysis Report Page 111
12
17

ITER G 73 DDD 2 01-06-06 W0.1
5.6 Test Blanket Modules in a Mid-Plane Port
Shielding properties of Test Blanket Modules (TBMs) in the equatorial Port 2 were studied in
the 3-D geometry by Monte Carlo method (See reference 1 ) to ensure that the TBMs don not
adversely affect the radiation condition in the VV/cryostat space.
Two basic TBMs concepts, developing by the ITER parties and adopted by the Test Blanket
Working Group-9, were presented in the calculational model (Figures 5.6-1, 5.6-2): the
helium-cooled Li4SiO4 ceramic breeder and the liquid lithium self-cooled concepts. In both
cases Beryllium is as a neutron multiplier.
Figure 5.6-1 TBMs Horizontal Cross- Section
1 G. E. Shatalov, A. A. Borisov, I. A. Kartashev, A. G. Serikov, S. V. Sheludyakov, O. L. Schipakin,
RF Design Support Contract: Neutronic Analysis of the ITER Vacuum Vessel / Cryostat Environment. Report
of RF HT for the First Quarter 2001, JF 04-01/1. April 2001, Moscow.
Nuclear Analysis Report Page 112

ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.6-2 Vertical Cross-Sections of the Ceramic He-cooled
Test Blanket Module (left) and Lithium Test Blanket Module (right)
In accordance with the remote handling and mounting requirements, the TBMs are installed
in the support frame of the test port. Behind the test modules, there is also an additional
shielding plug, designed to diminish radiation streaming effects from the TBMs and to reduce
the nuclear responses in superconducting coils and the cryostat. Impacts from other radiation
sources in the upper and divertor ports were minimized by thick shielding inserts in those
ports.
The prompt neutron and gamma-ray fluxes were calculated at many locations along the TFC,
PFC and the cryostat surfaces and in the toroidal detectors near the ports (See Figure 5.6-3).
These ccalculations were carried out assuming the nominal fusion power of 500 MW.
Then using the three steps MCNP-FISPACT-MCNP procedure 1 the gamma fields from
neutron activation of ex-vessel and structural components and dose rates were estimated in
locations around the port extension. This provided information on access conditions.
1 G. E. Shatalov, A. A. Borisov, I. A. Kartashev, A. G. Serikov, S. V. Sheludyakov, O. L. Schipakin, Interface
between MCNP and FISPACT codes for Dose Rate Estimation after Reactor Shutdown. Part 2 in: RF Design
Support Contract: Neutronic Analysis of the ITER Vacuum Vessel / Cryostat Environment. Report of RF HT
for the First Quarter 2001, JF 04-00/1. April 2000, Moscow.
Nuclear Analysis Report Page 113

ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.6-3 Total Neutron Flux (left), 10 7 cm -2 s -1 , and the 10-days Dose Rates (right),
μSv/h, in the Test Blanket Port Surrounding
The first 10-years testing period, corresponding to the first wall neutron fluence of about 0.1
MWa/m 2 was assumed in this consideration. The analysis performed so far has confirmed the
expectation of small radiation impact form the test modules, except in a few areas. Three
times higher total neutron flux was found in the local area at the port door near the cryostat in
comparison with the steel/water shielding blanket and plug in the port. The peak total neutron
flux of ~ 8 x 10 7 cm -2 ⋅s -1 was found at the upper intercoil structure under the port (See Figure
5.6-3).
It is shown, nevertheless, that the 10-days residual dose rates caused mainly by the activated
port walls and the side cryostat are still acceptable, ~ 40 _Sv/h and 90 _Sv/h , respectively,
at the cryostat flange, directly opposite the port, and at the port door.
An additional TFC heating from this TBMs port is small, ~ 10 W, i.e., within a statistical
error for a such kind of calculations.
However, a further work is required to verify local nuclear responses for the test module
structure to be developed in full details. Whenever the dose rates are not consistent with the
guidelines, design improvements of the TBMs should be made to reduce the exposure levels
by shielding.
Nuclear Analysis Report Page 114

ITER G 73 DDD 2 01-06-06 W0.1
5.7 Blanket Maintenance Port 1
There are four ports for blanket module maintenance purposes (#3, 12, 8 and 17). During
operation, plugs are inserted in all of those ports. Two of them have limiters with alignment
adjusting mechanism in their plug. The other two contain diagnostics.
This section describes the results of nuclear analysis around the blanket maintenance port
with limiters. Main radiation streaming path for this plug can be an annulus gap around the
limiter alignment support as well as 2 cm gap between the plug and port wall, which exists in
all standard ports. Nuclear heating rates, insulator dose rate in the magnet system and dose
rates around the port were evaluated.
The calculation for the ports with diagnostic plugs is described in Chapter 6.
5.7.1 3-D analysis model and the method
Figure 5.7-1 shows a maintenance port 3-D model visualised by the SABRINA code with
MCNP geometry input data. The geometry data for maintenance port plug produced by the
Japanese home team was inserted into the ITER basic model as of 26 April 2000. The port
extension cryostat, bioshield and inter-coil structures are also added to the basic model.
As shown in Figure 5.7-2 an annulus gap of 15 cm width locates behind the limiter module
around its alignment support. Since the radiation streaming through this gap into plug inner
space can be significant, rather thick plug frame ( 20 cm) is provided to prevent radiation
leaking from the inner space to the outside the plug.
Figure 5.7-1 3-D maintenance port model for MCNP (by SABRINA)
1 S. Sato, Y. Ohara and M. Akiba, “Radiation streaming analysis through equatorial port”, JP HT report of
D469-JA, 2001, June
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ITER G 73 DDD 2 01-06-06 W0.1
Blanket
gap: 2cm
Vacuum vessel
TF coil
gap: 15cm
Shield pug
20cm 20cm
with limiter
Limiter aligment support sysytem
PF coil support structure
10cm
Maintenance port
Figure 5.7-2 Horizontal cross section of the maintenance port model
Figure 5.7-3 Vertical cross section of the maintenance port model
5.7.2 Results of the analysis
1.1.1.1.1.1.1.1.2 A
5.7.2.1 Nuclear heating and insulator dose in the magnet system
Nuclear heating in the magnet system around the port is given in Table 5.7-1 and 5.7-2 and is
quite small giving no significant problem on the magnet system. Table 5.7-3 shows insulator
dose in the TF coil assuming 0.5 MW/m 2 averaged over entire first wall surface. The
insulator specific weight of 1.699 g/cm 3 is assumed. The absorbed energy by the insulator is
much less than the limiting value of 10 7 Gy.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 5.7-1 Nuclear heating in the TF coil around the blanket maintenance port
Coil case Insulator Winding Sub total Intercoil Total
Pack
structure
total(W) 0.594 1.05E-02 0.280 0.884 1.83 2.71
Table 5.7-2 Nuclear heating in the PF coil around the blanket maintenance port
PF#3 PF#4 Support
Structure
Total
Total (W) 7.54E-02 8.25E-03 3.07E-02 0.114
Table 5.7-3 Insulator dose in the TF coil around the blanket maintenance port
Inner side Insulator Side wall Insulator
Total (Gy) 5.05E+03 2.00E+03
5.7.2.2 Neutron fluxes and dose rate after shutdown around the port
Figures 5.7-4 and 5.7-5 show fast neutron fluxes (> 1MeV) and dose rate 106 second after
shutdown at the equatorial plane outside the port extension. Dose rates are quite low in
comparison with the design limit of 100 μSv/h. When we assume a similarly leaky plug on
the neighbouring port, the dose rate values should become larger by factor 2 since complete
shielding was assumed in the neighbouring port in this analysis. However, the expected dose
rate level is still lower than the limit with significant margin.
Figure 5.7-6 shows the ratio of the dose rate after shutdown and fast neutron flux during
operation. The values of the ratio scatters around the value of 2 x 10-5 μSv/h/(n/cm2/s), This
value can be used for quick estimation of the dose rate at ~11days after shutdown from the
fast neutron flux for the port with similar configuration.
PF coil support structure
TF coil
Maintenance port
3.4
Cryostat
5.8 6.8 6.5
4.0
5.5 5.4
7.5
6.0
5.8
unit: x 1E+5 n/cm2
fsd: 0.0881 - 0.1350
Figure 5.7-4 Fast (> 1 MeV) neutron flux distribution around the maintenance port
power:
Nuclear Analysis Report Page 117

ITER G 73 DDD 2 01-06-06 W0.1
PF coil support structure
TF coil
Maintenance port
9.48(6.58)
10.7(6.70)
7.01(6.51)
Cryostat
8.10(6.45)
11.3(6.77)
8.33(6.64)
9.05(7.16)
8.65(7.18)
8.45(6.45)
unit: μSv/hour
(fsd: %)
8.39(7.26)
Figure 5.7-5 Decay gamma-ray dose rates at 10 6 seconds after shutdown by 1-step
Monte Carlo method
PF coil support structure
TF coil
Maintenance port
2.37
1.84
2.06
1.51
Cryostat
1.41
1.52
1.23
unit: x 10-5μSv/hour/(1 neutron/cm2/s)
Figure 5.7-6 Ratio of fast (> 1 MeV) neutron flux and decay gamma-ray dose rates
One of the most probable locations near this port for personnel to access for maintenance
work is the space inside the port extension behind the plug. Figures 5.7-7, -8 and -9 show
dose rates after shutdown, fast neutron fluxes during operation and ratios of those two inside
the port extension. Again the level of the doses is low enough to allow personnel access to
this location.
Nuclear Analysis Report Page 118
A
1.58
1.29
1.55

ITER G 73 DDD 2 01-06-06 W0.1
9.34(5.24)
13.1(5.29)
14.8(6.14)
7.38(4.97)
7.16(5.84) 7.52(4.97)
9.86(4.98)
8.54(5.12)
6.90(4.60)
10.7(5.33)
8.63(5.04) 7.55(4.80)
unit: μSv/hour
(fsd)
7.60(4.67)
7.15(5.04)
7.31(4.80)
8.08(4.90)
Figure 5.7-7 Decay gamma-ray dose rates at 10 6 seconds after shutdown by one-step
Monte Carlo method
7.4(5.6)
12(6.1)
15(6.6)
5.7(7.4) 5.9(5.6) 5.8(5.5)
9.3(5.5)
7.6(5.2)
5.4(6.1)
10(5.8)
8.1(5.8) 6.8(5.9)
unit: x 1E+5 n/cm2
(fsd: %)
6.8(5.5)
5.9(5.8)
6.1(5.6)
7.4(5.5)
Figure 5.7-8 Fast (> 1 MeV) neutron flux distribution inside the port extension
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ITER G 73 DDD 2 01-06-06 W0.1
1.1
1.3
1.0 1.1
1.1
1.3 1.3
1.1
unit: x 1E-5 μSv/hour/(n/cm2)
1.1 1.1
Figure 5.7-9 Ratio of fast (> 1 MeV) neutron flux and decay gamma-ray dose rates
The values of the ratio of dose rate and fast neutron flux inside the port extension are almost
half of those outside the port. This implies that isotopes which are created by low energy
neutron, for example Co-60 and Fe-59, become more dominant outside the port.
Another location where personnel should access is the space for accessing PF coil #3 or #4
for their maintenance which locates above or below the equatorial plane. Figures 5.7- 10,-11,-
12 and -13 show the dose rates at those locations. The values of dose rate at those locations
are quite low in comparison with the design limit of 100 micro Sv/h.
PF coil(#3)
9.69(10.1)
18.1(9.49)
23.4(6.49)
23.8(6.21)
6.60(12.6)
Nuclear Analysis Report Page 120
1.3
1.1
7.19(14.3)
1.3
1.1
6.01(15.7)
1.1
1.2
unit: μSv/hour
(fsd: %)
1.47(30.7)
2.27(29.4)
Cryostat
2.93(33.6)
6.34(13.1)
6.66(15.0)
Biological shield
19.0(6.79)
5.87(15.6)
Port wall
Figure 5.7-10 Decay gamma-ray dose rates at 1E+6 seconds after shutdown above the
maintenance port ( center of the port)

ITER G 73 DDD 2 01-06-06 W0.1
15.3(12.3)
TF coil
12.8(10.0)
13.6(8.86)
PF coil support structure
PF coil(#3)
6.86(18.6)
7.31(11.7)
5.47(16.6)
5.37(14.8)
6.81(17.6)
4.92(14.5)
Cryostat
5.12(15.4)
unit: μSv/hour
(fsd: %)
1.45(33.6)
1.78(33.4)
1.98(29.5)
Biological shield
Figure 5.7-11 Decay gamma-ray dose rates at 1E+6 seconds after shutdown around the
maintenance port ( between the port)
Maintenance port
38.3(8.32)
27.1(8.08)
20.0(7.65)
20.2(7.98)
6.56(15.2)
10.0(12.2)
7.10(14.6)
2.33(34.6)
5.17(12.7)
Cryostat
unit: μSv/hour
(fsd: %)
Biological shield
0.848(27.2)
2.93(35.0)
PF coil(#4) 8.74(8.88)
8.27(12.1)
8.29(14.9)
7.54(13.9)
Figure 5.7-12 Decay gamma-ray dose rates at 10 6 seconds after shutdown below the
maintenance port ( center of the port)
Nuclear Analysis Report Page 121

ITER G 73 DDD 2 01-06-06 W0.1
TF coil
PF coil(#3)
7.12(16.2)
PF coil support structure
PF coil support structure
8.53(18.0)
6.85(16.9)
6.27(16.1)
6.92(13.6) 6.21(16.0) 6.52(18.8)
Cryostat
4.83(15.7)
unit: μSv/hour
(fsd: %)
1.56(25.4)
0.987(31.5)
1.28(25.3)
Biological shield
Figure 5.7-13 Decay gamma-ray dose rates at 10 6 seconds after shutdown below the
maintenance port ( between the port)
5.7.2.3 Dose rate as a function of time after shutdown
Time evolution of dose rate is shown in Figure 5.7-14. Although the figure shows that at a
typical location (“A” shown in Figure 5.7-4), it is generally applicable for all positions
around this port. Up to one day Mn-56 ( T1/2=2.58 h) contribution dominates. Then Co-58
(T1/2=70.8d) follows.
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ITER G 73 DDD 2 01-06-06 W0.1
1.E+03
1.E+02
1.E+01
1.E+00
1.E-01
1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 1.E+08
Time after Shutdown (sec)
ALL- 0
CR - 51
MN - 54
MN - 56
FE - 59
CO - 58
CO - 60
Figure 5.7-14 Time dependent dose rate behind the plug (point A in Figure 5.7-4)
5.7.2.4 Conclusion
The blanket maintenance port is well shielded and dose rate at 10 6 sec (~ 11days) around the
port is well below the limit (100 micro Sv/h) There will be no significant problem caused by
these ports for personnel access for the preparatory work involved in the removal or reinstallation
of those plugs.
5.8 Radiation conditions inside the divertor ports
5.8.1 Introduction
One of the main issues from the shielding point of view is to control the radiation streaming
through the major penetrations of the machine. The major concern is the activation generated
inside the cryostat. Workers during periodic maintenance would be subjected to high dose
rates, with restrictions on the access or to the time of work. The large openings in the
Vacuum Vessel, as the ports, weaken the shielding efficiency. Nevertheless they are required
for several operating functions.
At the level of the divertor there are 18 ports, one per sector, 10 of them are Cryo-Pump Ports
(CPP), to maintain the vacuum inside the VV, 3 are Remote Handling Ports (RHP) necessary
to remove the divertor cassette, the remaining 5 are used for diagnostics (DP). Each of these
three kinds of ports has its peculiarity, and problematic from the shielding point of view. It
can be assumed that the diagnostic ports will be provided with sufficient shield, plugs or
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ITER G 73 DDD 2 01-06-06 W0.1
similar, so that the leakage outside towards the cryostat will be minimal. For the RHP and the
CPP instead the solution is not in principle easy, because empty space inside the port is
intrinsically required.
A 3D specific neutronic analysis 1, 2 has been carried out to investigate what are the radiation
fields inside the divertor port area, if the shielding requirements in the cryostat are fulfilled
and if not which are the weak points to then suggest a proper modification. In the next
paragraphs the RHP has been analyzed, then the CPP will be considered.
5.8.2 Model description
To carry out this analysis a model has been used 3 with 20˚ symmetry with two halves ports,
one half is considered as RHP, and the other CPP. According to the model the number of
RHP and CPP is 9 each, not the actual one. But if mutual interactions of the ports are
negligible, the model can be used to make shielding studies for the two kinds of ports
independently.
5.8.3 Results for the RHP
Thc total neutron flux inside the divertor port when no action is taken to reduce the
opening of the port is about 2x10 11 (n/cm 2 s) in the front part of the port. There is a peak of
2x10 12 (n/cm 2 s) at the bottom entrance just below the cassette and a value of 9x10 11 (n/cm 2 s)
behind the gap between the divertor cassette and the blanket support (Figure 5.8.3-1). Most
of the streaming (50%) comes from that penetration, as it has been verified. A 46% is due to
the pumping slots in between the cassettes (Figure 5.8.3-2), while the contribution due to the
10 mm gaps in between the cassettes is only 3%.
1 G. Ruvutuso, H. Iida L. Petrizzi NAG-164-08-07-00 Radiation conditions inside the divertor ports Sept 2000
2 Neutronic Analysis of the ITER Vacuum Vessel/ Cryostat Environment, Report of the RF HT for the 3 rd
Quarter 2000, September 2000
3 G. Ruvutuso and H. Iida, NAG-159-08-06-00, “Three-Dimensional model of the ITER-FEAT reactor for
Monte Carlo nuclear analyses with MCNP”.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.8.3-1 3D view of module gaps , cassette gaps, and void space between the
cassette and the outboard blanket
Figure 5.8.3-2 3D view of pumping slots and cassette gaps
The dose rate has been estimated in different positions inside and around the divertor ports
(Figure 5.8.3-3 and 4), multiplying the neutron flux with energy > 1 MeV, by a factor 1.0x10 -
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ITER G 73 DDD 2 01-06-06 W0.1
5 μSv/h / [neutrons/(cm 2 s)] 1 . The values of the dose rate obtained, in the 5 positions
numbered in Figure 5.8.3-3 and 4, are summarized in Table 5.8.3-1, in the reference
configuration with the RHP port open.
.
Table 5.8.3-1 Dose rate values 10 6 s after shutdown [μSv/h] in the numbered positions
in the RHP
2
2
1 41.2
2 1.39x10 3
3 1.01x10 3
4 1.42x10 3
5 9.73x10 4
Figure 5.8.3-3 Poloidal section through the remote handling port with 3 scoring
positions.
1 H. Iida: NAG-166 Summary of calculation results for the values of the factors, which are required for
shutdown dose estimation Supplement to the NAG-157, Sept. 2000.
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3
1

ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.8.3-4 Toroidal section at the level of the remote handling port with scoring
positions.
In table 5.8.3-2 the values of the doses, in the reference case in which the port is open, are
compared with values obtained in the same places but with the RHP completely closed or
alternatively with an increased cassette body thickness. The solution “RHP completely
closed” means a solution in which the VV opening is filled with shielding material as it was a
poloidal continuation of the VV layer.
Table 5.8.3-2 Dose rate 10 6 s after shutdown [μSv/h] for different configurations
Position
Reference remote
handling ports
completely
closed
Thickness of
cassette body
increased by 7
cm
1 41.2 n. a. 57.6
2 1.39x10 3
n. a. 3.83x10 3
3 1.01x10 3
3.7 8.34x10 2
4 1.42x10 3
48.1 1.02x10 2
5 9.73x10 4
61.3 9.35x10 4
From the obtained results it can be stated that the radiation field inside the RHP is mainly due
to streaming through the pumping slot in between the cassettes and in gap between the
cassette and the bottom blanket support. An efficient shielding to the streaming cannot be
obtained simply increasing the divertor cassette thickness, but closing completely the RHP.
This solution can be easily met, because after the installation of the cassettes there is no need
to keep the port completely open. Viable solutions can foresee semi-permanent blocks of
shield that can close the mouth of the VV after the cassettes have been introduced, and that
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5

ITER G 73 DDD 2 01-06-06 W0.1
can be taken out again for the cassette maintenance. That solution assures a dose rate after
shutdown quite low with a sufficient margin.
5.8.4 Results for the CPP
As stated before, ten of the divertor ports are devoted to the pumping Cryo-Pump Port (CPP).
The problem of the neutron streaming through CPP ports look to be the most concerning,
because a minimum opening in the VV and a low impedance for the pump are an inherent
need.
On the other side the requirement is to reduce the radiation level in such a way that the
induced dose rate inside the cryostat facing the CPP is < 100 μSv/h, 12 days after shutdown
to let the workers operate for the due time. The nuclear heating on the cryo-pump itself
should not exceed 10 Watts (per pump, only in the 4 K˚ “array” which is one part of the
pump). In some of the CPPs an in viewing system (IVV) is installed, essentially pure silicon
probe with the purpose to control the plasma performances. To keep good performances (for
visible light) the probe should not be exposed to a radiation field such that the dpa is >0.5 and
the dose >10 10 Gy.
The basic model of ITER machine has been developed to include all the components related
to the pump, the extension port and the cryostat (Figures 5.8.4-1, and 2). The IVV probe
should pass through a channel, at the bottom of the cryo-pump to access the plasma region
that crosses a steel flange sustaining the pump itself. The In Viewing Channel (IVVC) has
cylindrical shape. After the installation of the probe it can be filled with some shielding
material to reduce the radiation streaming in the outer region.
Some spherical void regions have been defined for tallying purposes inside the port region to
study the variations of the response functions with the position. Positions 4 and 5 (Figures
5.8.4-3, 4) are facing the cryostat from the inside, most likely to be the place where the
workers will access.
Dose rates have been calculated using a new methodology, the so-called “one-step” method 1 .
Time adjusting decay factors have been taken from 2 , in which the M-GDR1 irradiation
conditions were assumed (0.3 MW y/m 2 in 20 years). Neutron fluxes have also been
calculated in the same positions, grouped in three wide energy bins: E < 0.1 MeV; 0.1 MeV<
E < 1. MeV; E > 1 MeV.
Different shielding configurations have been considered to see the impact on the response
functions. The reference port configuration foresees an extension of the vacuum vessel till the
rail sustaining the divertor cassette (“port closed”). This additional shield reduces the poloidal
mouth entrance of the port itself, without great impedance. The IVV channel has been
considered void in some cases (IVVC open) or closed by a plug (IVVC closed). Values are in
Table 5.8.4-1 and 2. In the table not all the figures are completed for tallying region number
5, but nevertheless it can be seen that a viable solution close to cryostat can be achieved when
1 L. Petrizzi, D. Valenza, H. Iida: Further development of a method of calculating the dose rate by means of
MCNP. NAG 143 13 12 99, Dec 1999
2 H. Iida: NAG-166 Summary of calculation results for the values of the factors, which are required for
shutdown dose estimation Supplement to the NAG-157, Sept. 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
all the penetrations are reduced. Borated steel added to wall ports can help, subtracting low
energy neutrons to the system.
Table 5.8.4-1 Dose rate values [μSv/h] calculated in 5 positions inside the CP Port for
different shielding configurations. The last column refers to a configuration with all the
penetrations open but with borate steel in the walls covering the walls of the port.
Open port
IVVC open
1 7.74 10 4
2 2.05 10 4
3 1.3310 4
4 3.21 10 3
5 9.01 10 2
Open port
IVVC closed
8.11 10 4
3.37 10 4
1.50 10 4
6.78 10 2
7.13 10 2
Closed port
IVVC open
6.12 10 4
9.43 10 3
7.11 10 3
2.77 10 3
Closed port
IVVC closed
(reference)
5.82 10 4
1.14 10 4
8.73 10 3
2.27 10 2
Open port
IVVC open,
borate steel in
the walls
7.73 10 4
1.65 10 4
8.60 10 3
1.07 10 3
Flux to dose conversion factor has been calculated as ratio of the neutron flux with energy
above 1 MeV and the dose rate, calculated by means of the direct “one step method”. Values
ranging between 0.5 and 3 10 5 , have been derived depending on the position. The geometry
with such complicated penetrations does not allow a unique factor to be derived. The neutron
energy spectrum is extremely variable, in the same way the 60 Co and 58 Co production, which
affects the dose rate 10 6 seconds after shutdown.
The requirement of having a dose rate < 100 μSv/h inside the cryostat 12 days after
shutdown, in the region of the CPP, can be almost achieved only in the reference design,
which foresees a partial closure of the CPP. Moreover, all the other possible penetrations in
the other components at valley of the CPP mouth should be closed. The VV extension added
at partial closure of the CPP, gives a reduction of about a factor 2-3 of the dose rate,
compared to the configuration in which no additional shield is included.
The nuclear heating on the cryo-pump has been calculated as well. The nuclear heating in the
cryo-pump has been calculated for two main shielding configurations: the cryo-pump port
“open”, and the same “closed”. The total nuclear heating on the pump with the CPP open is
420 W per pump, of which 4.6 W on the 4 K˚ “array” alone. The values become 157 W and
1.8 W respectively when the CPP is closed.
The nuclear damage has been calculated in positions where the IVV is supposed to be placed.
A maximum dpa of 3x10 -3 has been calculated assuming the IVV probe made of pure Si. The
peak value is calculated close to the divertor. The correspondingly accumulated dose is 2x10 9
Gy for a reference fluence of 0.3 MWy/m 2 . The IVV probe in Si does not look to have
problems of degradation due to neutron irradiation. Dpa are two order of magnitude lower
than the limit 90.5 dpa) and the peak-accumulated dose is 5 times lower than the limit (10 10
Gy).
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.8.4-1 Detail of the cryo-pump port. Configuration with partial closure of the
port and the IVVC (In Viewing Vessel Channel) open. SABRINA picture.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.8.4-2Detail of the cryo-pump system (SABRINA).
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 5.8.4-3 Poloidal section of the Cryo-Pump Port, with 4 of the scoring regions.
Configuration with port partially “closed” and IVVC open.
Nuclear Analysis Report Page 132
1
2 3 4
IVVC

ITER G 73 DDD 2 01-06-06 W0.1
1
4 K˚
array
1
Figure 5.8.4-4 Poloidal section of the Cryo-Pump Port, with the 5 scoring regions.
Configuration with port open and IVVC open.
Nuclear Analysis Report Page 133
2
2
2
3
3
3
4
5
4
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ITER G 73 DDD 2 01-06-06 W0.1
6 Radiation Properties of Diagnostic System Plugs
A number of diagnostic plug layouts have been and will be analysed to cover all developed
detailed diagnostic systems. Some of them do not alter the bulk shielding efficiency. In other
cases when diagnostic access apertures affect the effective blanket / vacuum vessel shielding
capability, this is recovered by labyrinthine access penetrations in special steel/water
shielding plugs. Several representative configurations are shown in this chapter.
6.1 Vertical Neutron Camera
Neutron cameras with horizontal and vertical views have been designed for ITER 1 to
provide line-integral measurements of the D-T neutron emissivity along the available sight
lines. The sight lines view the ITER plasma through slots in the shielding blanket and
penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows.
Their spatial sampling is adequate for satisfying measurement requirements. But
simultaneously, the diagnostic system has to satisfy also nuclear shielding and maintenance
requirements. The expected nuclear performance of the vertical neutron camera arrangement
for ITER is described below.
6.1.1 Preliminary analysis and design consideration
The radiation conditions and nuclear performance of different diagnostic systems for ITER
were considered in reference 2 , including a "short stub" neutron camera design proposed in
reference 3 . A fan-shaped arrays of collimated flight tubes, with suitably chosen detectors
situated outside the cryostat or the bio-shield, allowing access to detectors were foreseen in
this concept. Even a limited field of view required a cryostat interface with a significant
radial extension.
A massive radiation shield surrounded flight tubes and detectors in order to provide
collimating of neutrons along each sight line and to prevent nuclear activation of
neighbouring components. It imposed load constraints on the design of supporting structures.
A neutron streaming analysis of the "short stub" neutron camera concept through slots, gaps
and channels introduced by this system into the bulk radiation shield was performed as a
basis for a further nuclear response evaluation in the system surrounding. Simplified methods
and tools as well as the results of the comprehensive nuclear analysis reference 4 carried out
for the 1998 ITER design have been used for the analysis.
1 L. C. Johnson, C. W. Barnes, P. Batistoni, C. Fiore, G. Janeschitz, V. Khripunov, A. Krasilnikov, F. B.
Marcus, T. Nishitani, G. Sadler, M. Sasao, V. Zaveriaev, and the ITER Joint Central Team and Home Teams,
Analysis of Neutron Cameras for ITER. Review of Scientific Instruments, Vol. 70, No. 1, pp 1145-1148, Jan.
1999.
2 V. Khripunov, Preliminary Analysis of the Nuclear Environment for LAM and IAM Diagnostic Systems.
G 55 RI 8 99-02-09 W 0.1 (NAG-123-09-02-99).
3 Ph. Edmonds, A Concept for the “Short Stub” Neutron Camera Design for IAM. Garching, 19 January 1999.
4 R. T. Santoro, V. Khripunov, H. Iida et al., ITER Nuclear Analysis Report. G 73 DDD 1 98-06-17 W0.2,
NAG-101-98-06-17-CDR, June 1998, Garching.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 6.1-1 Vertical and Horizontal Neutron Cameras Layouts
The study showed that an impact of the neutron camera on the important nuclear responses in
the vacuum vessel/TFC/top cryostat region including the operational parameters and the
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ITER G 73 DDD 2 01-06-06 W0.1
residual activity should be low enough in comparison with the radiation conditions provided
by the bulk radiation shield.
It was shown also that optimisation of shielding design is possible. In particular, a very long
plug length ~1.8 m in the cryostat top direction increasing the total weight of the neutron
camera block is not desirable from the shielding point of view. A ~30-cm 70%SS / 30%H2O
shield was recommended as beam dump, and also a ~10-cm annular shield around the
channels, to prevent a correlation between radiation fields inside and outside detectors in the
channels. Besides, it was recommended to remove an intermediate massive shielding around
the collimators up to the cryostat lid.
6.1.2 Model
In the arrangement of the current neutron camera concept 1 without inter-space boundary
between the plasma and the cryostat (See Figure 6.1-1), the vertical part is a six-chord
system. Details of this system are shown on the right-hand side in Figure 6.1-1.
It reflect the idea to use the existing gaps between blanket modules and the intercoil
structures to collimate D-T neutrons without introduction of additional structural elements for
that purpose. The concept is consistent with the present standard blanket segmentation: the
neutron paths have neither interference with the blanket modules nor filler shield modules,
but those only go through the upper port plugs. Two rows of penetration holes in each port
will be possible through the diagnostic plug and the vacuum vessel. Each row has 2-3
penetrations with about 300 mm interval.
Shield is necessary above the penetration holes in the vacuum vessel to prevent the fast
neutron flux from spreading. Shielding blocks with larger holes are considered to attach to
the outer inter coil structure.
Second collimators (annular shields) are hung down from the cryostat to catch the straight
and forward scattering neutrons. Neutron detectors are mounted on the top cryostat. These
shielding blocks and beam dump are thick enough to allow hands-on approach to the cryostat
floor.
6.1.3 Nuclear performance
The analysis of the proposed model was carried out, commented on the model to be suitable
from the view point of nuclear requirements. It included available results of a 2-D modelling
of the channels 2 , a 3-D modelling of the upper part of the reactor 3 (without a vertical neutron
camera), and the following superimposing of the streaming effects.
1 K. Ebisawa, Current Design Status of the Vertical Neutron Camera. Interoffice Memorandum.
ITER Naka JWS, 16 May , 2000.
2 V. Khripunov, Draft Notes on Vertical Neutron Camera. Interoffice Memorandum. ITER Garching JWS, 30
August, 2000.
3 Neutronic Analysis of the ITER Vacuum Vessel/Cryostat Environment. Report of RF HT for the 2-nd quarter
2000. JF-04-00/2, July 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
6.1.3.1 Specific energy release
The specific nuclear energy deposition in the channel walls was estimated at different
locations. It is about ~ 0.01 W/cc and ~ 0.002 W/cc in the first steel window (behind the
blanket) and second windows, respectively. In the intercoil structure region it is much lower,
~ 4 10 -5 W/cc, and the total additional heating of the annular channel shielding here is small,
~ 0.3 W. Nuclear energy deposition in the surface layer of the cryostat top, the annular shield
and wall outside the detector volume is about ~2 10 -6 W/cc. A nuclear heating ~3 10 -5 W/cc is
expected in steel element inside the detector volume due to neutron streaming through a
channel.
Thus, nuclear heating of the channel walls and the surrounding shield located near the
intercoil structures is small, about a fraction of Watt per a channel. It will not impact
essentially on the required cryogenic power.
6.1.3.2 He-production in steel channel walls
The estimated He-production in steel walls of a separate 40-mm collimator channel in the
vacuum vessel region does not exceed 0.05 He appm at the end of the DT operational period
(the first wall neutron fluence of ~ 0.3 MWa/m 2 ). It is smaller than the re-welding limit of ~1
He appm. Thus, re-welding of the vacuum vessel and windows should be possible. However
these conclusion has to be confirmed for a more developed neutron camera design taking into
account the water regions nearby and possible interference between neighbour channels.
6.1.3.3 Activation
The activation of the annular shield and the detector shield from outside, by the
“background” neutrons, is not problematic. The expected residual dose rate at these structures
two weeks after shutdown will not exceed the 100 µSv/hr limit for hands-on maintenance.
The collimated neutrons increase locally the surface activity and contact dose rates at the
annular shield flange and inside the detector volume by ~ 1 order of magnitude. However, the
expected average dose rate around these regions (not inside the detector volume) should be
acceptable and will not change essentially the radiation conditions below the neutron camera.
But absolute values should be recalculated in 3-D geometry.
6.1.4 Neutron and Photon Flux Distributions in the Camera Surrounding
6.1.4.1 “Background” Fluxes
The typical neutron and gamma-ray fluxes from the main neutron source in the plasma
chamber are given in Table 6.1-1 for different locations starting from the first wall of the
upper blanket module 9, where the neutron wall loading is ~0.43 MW/m 2 , along the sight
lines, and up to the upper cryostat. These 3-D “background” fluxes are not disturbed by the
neutron streaming through the camera penetrations.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 6.1-1 “Background” Fluxes from the First Wall to the Top Cryostat
at Different Locations of the Neutron Camera Elements
Flux component n-DT n-fast n-total gammatotal
Energy, MeV 14.1 > 0.1 > 0 > 0
Flux values, cm -2 s -1
First Wall 3.0 10 13
1.1 10 14
1.9 10 14
7.1 10 13
First Window in VV 4.4 10 10
Second Window in VV
VV / IC Structure gap 2.8 10 6
9 10 11
1.2 10 8
Annular Shield Exit 5.1 10 5
1.1 10 7
Detector Volume ~ 1 ~ 2 10 2
1.8 10 12
1.9 10 8
1.6 10 7
~ 5 10 2
1.8 10 12
1.0 10 8
1.1 10 7
~ 5 10 2
It is seen from this table that a DT- flux attenuation factor in the bulk shield (including the
blanket, VV and upper TFC part) is ~ 2 10 -8 . The geometry attenuation factor from the outer
VV surface to the top cryostat is ~ 2 times.
The “background” fast and DT-neutron flux components attenuate further by ~ 3 orders of
magnitude in the annular shield hung down from the cryostat and in the detector shield.
6.1.4.2 Collimated Flux Components
The collimated neutron flux components were calculated using a 2-D model of the vertical
neutron camera channels (Table 6.1-2).
Table 6.1-2 Main Collimated Flux Components
in the Detector Volume of the Central Channel
Flux component Energy, MeV Flux value, cm -2 s -1
n-DT 14.1 ~ 7 10 7
n-fast > 0.1 ~ 3 10 8
n-tot > 0 ~ 3 10 8
< 0.1 ~ 1.3 10 7
n-thermal < 0.4 10 -6
~ 15
Gamma-tot > 0 ~ 2 10 8
These components reflect a flux attenuation along the central diagnostic channel directed to
the plasma core. The collimated DT-flux attenuates by factor of ~3 10 -6 in the way from the
first wall to the detector region. In the detector region this flux component is by 2 - 3 orders
of magnitude higher than that inside the cryostat.
Other collimated fluxes are 1-2 orders of magnitude higher than the fluxes at the annular
shield exit near the top cryostat. But due to the annular shield effect this difference increases
to ~5 orders of magnitude.
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ITER G 73 DDD 2 01-06-06 W0.1
It should be noted, however, that according to our previous consideration 1 both the total and
partial collimated neutron fluxes in other, not the central channels which are directed to the
plasma periphery, due to different “linear integrals” may be about 10 times smaller. That is
why a correct (but very time consuming) 3-D modelling of the DT-neutron angular
distributions at the channel exits and their transport through the channels is still required.
6.1.4.3 Neutron spectra evolution along the channels
The “collimated” neutron flux components caused by neutron streaming through the channels
(30-50 mm dia, ~8.5 m long) should be superimposed on the “background” spectra. As a
result the evolution of the neutron spectrum from the first wall to the detector volume is
shown in Figure 6.1-2.
Fluxes, 1/(cm2s) per Lethargy
1.E+16
1.E+15
1.E+14
1.E+13
1.E+12
1.E+11
1.E+10
1.E+09
1.E+08
1.E+07
1.E+06
1.E+05
1.E+04
1.E+03
1.E+02
1.E+01
1.E+00
1.E-01
1.E-02
1.E-
01
"Background" and Collimated Spectra
1.E
+00
1.E
+01
1.E
+02
1.E
+03
1.E
+04
1.E
+05
Neutron Energy, eV
Nuclear Analysis Report Page 139
1.E
+06
1.E
+07
1.E
+08
FW
Bl/VV gap
VV/ICS-gap
at Annular Shield
Detector Volume
Collimated Fluxes
Figure 6.1-2 Evolution of the Neutron Spectrum
from the First Wall to the Detector Volume behind the Cryostat
1 V. Khripunov, Water Diagnostic Loop Parameters. G 72 RI 198-01-30 W 0.1, Garching, 5 January 1998.

ITER G 73 DDD 2 01-06-06 W0.1
6.1.5 3-D Modelling
The preliminary analysis was principally confirmed as a result of 3-D Monte Carlo
calculations 1 . Details of a simplified neutron camera model introduced in the basic ITER
model are shown in Figure 6.1-3.
Both the volumetric and surface D-T neutron sources were used in the plasma chamber and at
the collimator entrances to get accurate evaluation of neutron flux distributions, different
nuclear responses and activation of surrounding structures, starting from the first wall up to
the detectors and beam dump behind the upper cryostat lid.
It is seen from Table 6.1-3 that energy group fluxes are very similar in the channels near the
first wall. However, they differ by one order of magnitude at the detector locations in the
beam dump behind the cryostat depending on the D-T neutron currents in each channel
(channel inclinations relating to the plasma core).
Table 6.1-3 Neutron Flux Attenuation along the Channels
Energy,
D-T n-fast
MeV < 0.1 0.1 0.1
First Wall (channel openings)
Channels
1- 6 (avr) 8.5 10 13
7.8 10 13
3.0 10 13
1.1 10 14
Beam Dump (behind the Cryostat)
1 3.3 10 7
2.1 10 7
7.9 10 6
2.9 10 7
2 2.0 10 7
2.1 10 7
1.4 10 7
3.6 10 7
3 2.1 10 7
2.1 10 7
2.3 10 7
4.4 10 7
4 2.2 10 7
2.2 10 7
3.2 10 7
5.3 10 7
5 2.9 10 7
3.0 10 7
5.2 10 7
8.2 10 7
6 2.6 10 7
2.9 10 7
7.3 10 7
1.0 10 8
Total
> 0
1.9 10 14
6.2 10 7
5.6 10 7
6.5 10 7
7.5 10 7
1.1 10 8
1.3 10 8
The total neutron flux at the top cryostat due to neutron streaming through six channels is ~5-
8% higher than the background value. At the side cryostat wall this effect is lower, ~5-1 %.
Additional residual gamma-sources were identified in the Upper Outer Intercoil Structure
(UOIS, See Figure 6.1-3) which is activated by neutrons penetrating through the six neutron
camera channels. The dose rates 10-days after reactor shutdown at the upper cryostat surface
caused by these gamma-ray sources do not exceed 8 µSv/h.
1 G. E. Shatalov, A. A. Borisov, I. A. Kartashev, A. G. Serikov, S. V. Sheludyakov, O. L. Schipakin,
RF Design Support Contract: Neutronic Analysis of the ITER Vacuum Vessel / Cryostat Environment. Report
of RF HT for the First Quarter 2001, JF 04-01/1. April 2001, Moscow.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 6.1-3 Vertical Neutron Camera Model (upper)
and its Layout in the Upper Port and the Cryostat Regions (lower)
6.2 Edge Thomson Scattering System in Upper Port
A three dimensional 20° model of the edge Thomson scattering system (ETSS) in the upper
port, is shown in Figure 6.2-1.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure 6.2-1 A 3-D Model of the Edge Thomson Scattering System in the Upper Port
Using this model neutron flux and energy deposition in mirror materials and vacuum window
were calculated.
An appropriate choice of the biasing technique was required for the labyrinth analysis 1 . It
shows that major elements of the system do not give rise to unacceptable neutron fluxes and
heat deposition in the system and its surrounding.
The first mirror of the ETSS is located ~1 m behind the first wall in order to reduce the
radiation loads and neutron streaming to the flange plate. The calculated specific nuclear
heating here does not exceed ~ 20 mW/cm 3 , the fast neutron flux ~ 5.6 10 11 cm -2 s -1 and total
neutron flux ~ 1.2 10 12 cm -2 s -1 . Thus simple water-cooled, stainless steel mirrors can be used
here.
Heating in silica lenses (See Figure 6.2-1) is even much lower, below 1 _W/cm 3 . This is by
2-3 orders of magnitude lower than in the first mirror material.
Due to the “dog-leg” configuration the neutron and photon fluxes attenuate along the channel
by 5 orders of magnitude at the outlet lens. This does not change the average flux level inside
the cryostat from that expected with a solid plug.
Nuclear heat loads on the nearest cryogenic systems will not be disturbed by this diagnostic
plug.
Additional interaction analysis appears to be required if other diagnostic systems are inserted
in the port.
1 Neutronic Analysis of the ITER Diagnostic Systems. Report of RF HT for the 2 and 3-d quarters 2000. June
and September 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
6.3 LIDAR and Polarimetry Diagnostic Systems in the Integrated
Mid-Plane Port
Another representative example is the combined LIDAR and polarimetry diagnostic system
in the equatorial port (Figure 6.3-1) where the resulting streaming through the large channel
might be a matter of concern.
Initially a special 3-D model of LIDAR system in a shielding diagnostic plug in an equatorial
port was developed for neutron transport calculations 1 . Then the second polarimetry system
was added.
Figure 6.3-1 LIDAR and Polarimetry Diagnostic System Models for Nuclear Analysis
6.3.1 Plug model description
The 3-D model of the integrated diagnostic port plug includes steel frame which consists of
the 15-cm base and two 10-cm support plates (10 cm), attached to the 16-cm flange. The
space between support plates is filled with 60%SS/40% water shielding structure with the 20mm
gaps around.
The conic channel of the LIDAR (~18 cm id at the first wall) is located in one half of the port
model. It includes five 30-50-cm vanadium mirrors and two ~15-cm quartz windows in the
seal flange and in the cryostat.
In the second part of the port model a polarimetry system is represented by 20 cylindrical
channels (14 cm id) arranged in two rows (See also Figure 6.3-2). There are also two sets of
plane steel mirrors and primary quartz windows (dia 11.5-cm) in every channel.
An entrance of the channels in the first wall and blanket shielding block is represented by the
23 cm x14 cm rectangular slot.
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ITER G 73 DDD 2 01-06-06 W0.1
This model of the combined diagnostic plug was introduced then into the basic 3-D model of
ITER and the radiation transport calculations were carried out using the MCNP-4B code
system 1 . The radiation streaming through other ports were excluded from this consideration.
Figure 6.3-2 LIDAR and Polarimetry System Cross-Sections
The distribution of neutron flux was calculated from the first wall up to the seal flange along
the port plug and diagnostic channels, with heat deposition in mirrors, windows and in the
body of the diagnostic plug.
Some results of this modeling are presented below normalized to the nominal fusion power of
500 MW.
6.3.2 LIDAR
The neutron group flux distributions along the LIDAR dog-leg channel were calculated 2
separately for the LIDAR system and the combined system models (Table 6.3-1).
The nuclear heat deposition in the LIDAR mirrors and windows is given for these two cases
in Table 6.3-2.
It is seen from these tables that the polarimetry diagnostic system increases the neutron flux
and nuclear heating at the neighbor LIDAR elements. The flux at the LIDAR mirror 3 is up
from ~7x10 8 cm -2 s -1 to ~2.9x10 9 cm -2 s -1 with a nuclear heat deposition up from 12 μW/cm 3 to
100 μW/cm 3 . The flux at the LIDAR window is up from 2.1x10 7 cm -2 s -1 to 9.5x10 7 cm -2 s -1
with a nuclear heat deposition up from 0.14 μW/cm 3 to ~2 μW/cm 3 .
1 MCNP 4B, Monte Carlo N-Particle Transport System. Los Alamos National Laboratory, Los Alamos, New
Mexico. Ed. by J. Briesmeister, LA-12625-M, November, 1993.
2 Neutronic Analysis of the ITER Vacuum Vessel/Cryostat Environment. Report of RF HT for the 3-d and 4-th
quarters 2000. JF-04-00/3, JF-04-00/4. September, December 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 6.3-1 Neutron Fluxes in the LIDAR System Elements
(neutron wall loading 0.7 MW/m 2 , fusion power 500 MW)
Energy, MeV 13.5 > 0
(total)
Locations (See Figure 6.3-1) cm -2 s -1
cm -2 s -1
cm -2 s -1
cm -2 s -1
Mirror 1 2.7 10 11
2.6 10 11
1.3 10 11
6.5 10 11
Mirror 2 1.9 10 9
1.0 10 9
2.9 10 9
Mirror 3 5.7 10 8
1.6 10 8
7.3
Primary Vacuum Window 1 2.1
Shutter 1.9 10 10
Seal flange:
- LIDAR part 1.1 10 7
2.1 10 10
2.8 10 7
0.4 10 10
Combined
model
> 0 (total)
6 10 11
3.6 10 9
2.9 10 9
Nuclear Analysis Report Page 145
10 8
10 7
4.4
10 10
- 3.9
10 7
9.5 10 7
2 10 11
6.5 10 7
- polarimetry system
part
1.1 10 8
Bellows 1.1 10 7
Cryostat 1.5 10 7
Mirror 4 2.0 10 6
1.3 10 6
- 3.3
10 6
1.4 10 7
Mirror 5 2.5 10 6
1.6 10 6
4.1 1.5 10 7
Vacuum Window 2 1.9
10 6
Table 6.3-2 Specific Nuclear Heating in the LIDAR Elements
Locations
LIDAR Combined Units
(See Figure 6.3-1)
model model
Mirror 1 15 20 mW/cm 3
Mirror 2 50 70 _W/cm 3
Mirror 3 12 100 ”
Primary Vacuum Window 1 0.14 2
Shutter 5.3 8 mW/cm 3
Seal flange:
- LIDAR part - 1 _W/cm 3
- polarimetry system part 4 ”
Bellows 0.1 ”
Cryostat 0.2 ”
Mirror 4 0.03 0.2 ”
Mirror 5 0.02 0.2 ”
Vacuum Window 2 0.006 - ”
10 6

ITER G 73 DDD 2 01-06-06 W0.1
6.3.3 Polarimetry System
The maximum fast neutron (En>0.1 MeV) fluxes at the mirrors (Figure 6.3-2) are ~1-2 x10 12
cm -2 s -1 , and the total flux (En>0) is ~2 times higher.
Primary vacuum windows receive much lower neutron fluxes,

ITER G 73 DDD 2 01-06-06 W0.1
from mirror to mirror along the broken cavern axis on the way from the first wall to the port
periphery.
The following features of the MSE system important from radiation shielding view point
were identified.
A big cavern for light passage with the cross section increasing from ~ 30 cm x 60 cm in the
first wall region to 30 cm x 80 cm at the port exit removes an essential part of the outboard
radiation shield and represents a way for neutron streaming to the surrounding structures.
Thus, in spite of a “dog-leg” cavern configuration, the MSE is a leaky system like another
Diagnostic Neutron Beam system with the straight (~ 40 cm x 40 cm) channel and therefore
an adequate shielding is required.
The essential part of the cavern walls is facing the plasma. Thus all kinds of plasma/wall
interactions are possible (See Table 6.4-1) and a port plug design should be similar to the first
wall and shielding blanket design.
Table 6.4-1 Nuclear Parameters for the Central Outboard Part
of the First Wall
Nominal Fusion Power 500 MW
Neutron Wall Loading 0.76 MW/m 2
Effective (local) First Wall Neutron Fluence
(for the average neutron fluence of 0.3 MWa/m 2 ) ~ 0.41 MWa/m 2
DT (14.1 MeV)-neutron Flux 4.5 10 13 cm -2 s -1
Fast (> 0.1 MeV) neutron Flux ~ 1.9 10 14 cm -2 s -1
Gamma-ray Flux ~ 1.9 10 14 cm -2 s -1
Specific Nuclear (n+_) Energy Release in SS :
- 0 cm from the first wall (in the Be-layer position): ~ 8.1 W/cc
- 20 cm from the first wall (in the blanket body) ~ 0.7 W/cc
Surface Radiation Heating (maximum) ~ 19 W/cm 2
(It should be remembered that operational parameters estimated below could be ~40% higher,
and fluence dependent characteristics ~ 20-60% higher in case of the enhanced fusion power
~ 700 MW and the longer operational period, corresponding to the first wall neutron fluence
0.5 MWa/m 2 ).
It was reasonable to conclude that the nuclear heating in the TFC and intercoil structure from
radiation leaking through the port plug and walls is low.
However, the residual gamma-source intensity of the outer port extension surface caused by
this leakage was unacceptable for limited manual access to cryogenic system nearby.
Based on the results of this consideration the preliminary design of MSE system was
significantly modified to meet the radiation shielding requirements. That included: - a smaller
cavern cross section still acceptable for diagnostic purposes; - a changed angle of the cavern
axis between mirrors; - an additional ~ 20-30 cm steel/water radiation shield behind the
penetrated port closing plate to diminish cryostat activation, and others.
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ITER G 73 DDD 2 01-06-06 W0.1
6.4.1 3-D MSE Modelling
After these improvements to the MSE system a special 3-D model of this system in the upper
part of the equatorial port was developed and included into the basic 3-D ITER shown in
Figures 6.4-1, 6.4-2.
a) Plan view b) Cut plan
Figure 6.4-1 3-D Model of the MSE Diagnostic System in the Equatorial Port
It includes: - a 0.60 SS/0.40 H2O shielding plug with a dog-leg cavity, formed by five
channels of rectangular cross sections (an aperture ~ 30 x 40 cm); - rectangular (M1=50x30
cm, M2=55x10 cm) and circular (M3=50 cm dia, M4=60 cm dia) steel mirrors (3 cm in
thick); - a quartz vacuum window (W1=10 cm dia, 2 cm thick.), See Figure 6.4-1; - a 13-cm
steel frame; - 2-cm gaps between the frame, and - the port walls.
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ITER G 73 DDD 2 01-06-06 W0.1
a) Y=0 - a symmetry plane b) Y= -45 cm
c) Y= -45 cm - an additional shielding pug at the vacuum window
Figure 6.4-2 MSE Model Elevation View
An appropriate choice of the biasing technique ("splitting and Russian roulette"), additional
shielding plugs in the bottom part of the diagnostic port (Figure 6.4-2), and in other ports
were required for the labyrinth analysis 1 to reduce a variance of calculated fluxes and nuclear
responses and to distinguish a MSE contribution to the nuclear performance in the port
surrounding.
6.4.2 Nuclear Response Distributions along the Diagnostic Channel
Using this model the distributions of neutron and gamma-ray fluxes were calculated from the
first wall up to the seal flange along the port plug and diagnostic channel, with heat
deposition in mirrors, windows and in the body of the diagnostic plug (Figures 6.4-3, 6.4-4).
1 Neutronic Analysis of the ITER Vacuum Vessel/Cryostat Environment. Report of RF HT for the 1-st Quarter
2001. JF-04-01/1. March 2001.
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ITER G 73 DDD 2 01-06-06 W0.1
n/ cm ^ 2 / s
total neutron flu x,
1.0E+14
1.0E+13
1.0E+12
1.0E+11
1.0E+10
1.0E+9
1.0E+8
1.0E+7
chan.1
chan.2
support plate
front side back
chan.3
m1
m2
chan.4
800 900 1000 1100 1200
X coordinate along support plate, cm
Nuclear Analysis Report Page 150
m3
chan.5
m3
m4
flange
m4
flange
port extension
Figure 6.4-3 Total Neutron Flux Distribution along the MSE System Channel
W/ cm ^ 3
deposition,
volu metric energy
1.0E+1
1.0E+0
1.0E-1
1.0E-2
1.0E-3
1.0E-4
1.0E-5
1.0E-6
1.0E-7
chan.1
front
chan.2
chan.3
support plate
side
m1
chan.4
m2
chan.5
W1
W1
back
m3
m4
m3
m4
flange
flange
port extension
W1
W1
800 900 1000 1100 1200
X coordinate along support plate, cm
Figure 6.4-4 Specific Nuclear Energy Deposition in the Channel Walls and Mirrors

ITER G 73 DDD 2 01-06-06 W0.1
The calculated specific nuclear heating of the first steel mirror located ~ 1 m behind the first
wall does not exceed ~ 0.4 mW/cm 3 , and total neutron flux ~ 1.2 10 13 cm -2 s -1 (Table 6.4-2). It
is only about one order of magnitude lower than at the first wall (See Table 6.4-1) and ~3
times lower than for the MSE design considered preliminary in reference 1 .
Detector locations
(Figures 6.4-1, 6.4-2)
Table 6.4-2 Neutron Flux and Nuclear Energy Distributions
along the MSE System Channel (fusion power 500 MW)
fast (En>
0.1 MeV)
Neutron Fluxes, cm -2 s -1
total (En> 0) statistical
error, %
1.2 x 10 13
Nuclear Energy
Deposition
W
mW/cm 3
Mirror M1 7.5 x 10 12
3 420 1800
Mirror M2 1.5 x 10 11
3.4 x 10 11
9 11 16
Mirror M3 2.3 x 10 9
8.8 x 10 9
9 0.13 0.73
Mirror M4 5.6 x 10 8
1.6 x 10 9
10 0.038 0.33
2.2 x 10 8
4.0 x 10 8
Vacuum Window (W1)
*)
3.1 x 10
12 0.013 0.0018
8
4.3 x 10 8
18
0.4 x 10 8
(2.6-1.6) x
10 8
Flange
11 0.0007 -
*)
(0.3-0.2) x
10 8
(0.9-0.6) x
10 8
25
2.6 x 10 7
7.7 x 10 7
Cryostat
9 0.0002
*)
1.1 x 10 7
4.4 x 10 7
16
*)
With a shielding plug behind the closure port door shown in Figure 6.4-2c.
A surface heating of the steel mirror M1 from plasma irradiation through the first part of
channel ~0.2 W/cm 2 is expected based on the preliminary estimates reference 1 , and damage
of steel does not exceed ~0.01 dpa at the end of the D-T operational period (0.3 MWa/m 2 ).
Then, the neutron and photon fluxes, nuclear heat deposition and other nuclear responses
attenuate along the channel by 5 orders of magnitude at the outlet window.
The total neutron flux at the cryostat apart from the port door and the vacuum window W1 is
about ~10 8 cm -2 s -1 . That does not change the average flux level inside the cryostat from that
expected with a solid plug in the port and is acceptable to allow a limited access to the port
flange.
The residual gamma-ray transport modelling was performed directly for this region.
The residual dose rate between the seal flange and the cryostat due to neutron
streaming throughout the MSE system structure is ~ 60- 80 µSv/h two weeks after
shutdown during the reactor life time.
The analysis shows that the major part of the MSE system does not give rise neutron fluxes
and heat deposition in the port surrounding. Local nuclear heat loads on the nearest TFC are
only by 25-30% higher than in case of a very well shielded port.
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ITER G 73 DDD 2 01-06-06 W0.1
In case of using an additional annular shielding block at the outlet window W1 (See Figure
6.4-2c), the neutron fluxes at the TFC outer case and at the port door can be reduced by 5-
15% and ~50%, respectively.
The total nuclear heat deposition in the port structures is about 3.2 MW, from which ~1.4
MW are referred to the shielding blanket in front of the MSE diagnostic system, ~40 kW
release in the diagnostic plug itself, and the rest - in the shielding structure below the MSE
system (Figure 6.4-2).
Additional interaction analysis appears to be required for the integrating MSE together with
the NPA system located in the same port (instead of the temporary shielding structure used in
this model).
6.5 The In-Vessel Viewing System Study
As was mentioned in Section 5.8, the in-vessel viewing (IVV) system, suggested in
reference 1 for the first wall coverage inspection after reactor shutdown, uses the 22 cm x 21
cm penetrations between the central divertor cassettes and the blanket. It is integrated in six
evenly distributed divertor pumping ports taking into account viewing resolution and port
availability. An inclined trajectory inside the vessel is required to give adequate first wall
viewing. A calculational model of the system is shown in Figure 6.5-1.
Figure 6.5-1 IVV Channel in the Central Cassette of the Divertor Pumping Port
1 E. Martin, R. Tivey, G. Jantschitz, A. Antipenkov, et al, “ITER-FEAT Divertor Maintenance and Integration”,
the 21th SOFT Madrid, September 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
A multistep nuclear analysis was performed to integrate this IVV system at the divertor level
(See references 1 2 3 4 5 ). It shows that continuous shielding is required to prevent or, at least,
to diminish consequences of neutron streaming through these channels, such as enhanced
neutron fluxes and activation, and an additional nuclear heating of the cryo-pumps and
magnets.
That is why to avoid neutron streaming during reactor operation the penetration is closed at
the divertor rear surface by using a passively cooled gate type shield block, that is actuated by
a rod and a jack outside the bio-shield.
A shield plug is incorporated also to ensure the vessel port flange and the bio-shield
continuity during machine operation.
6.5.1 Streaming effects
Streaming effects of the IVV system were also studied in references 6 7 8 using the 3dimensional
code MCNP-4B and a 3-D model of ITER. The IVV penetration was
incorporated in the pumping port configuration as a 22-cm straight channel in the central
cassette (Figure 6.5-1).
Neutron transport modeling was performed assuming both the volume and surface neutron
sources at the penetration mouth having similar angular and energy distributions.
Besides, an additional shielding was proposed behind the outboard blanket modules (No. 17)
to diminish radiation streaming from the upper blanket. Thus radiation sources outside the
channel were excluded from the consideration.
The next neutron field parameters, normalized to the fusion power of 500 MW, were used in
this region as the starting values:
1 V. Khripunov, Preliminary Analysis of the Nuclear Environment for LAM and IAM Diagnostic Systems.
G 55 RI 8 99-02-09 W 0.1 (NAG-123-09-02-99). Garching, February, 1999.
2 E. Martin and V. Khripunov, In-Vessel Viewing and Metrology Systems. Design Progress report at the
Remote Handling Meeting. Garching, April 23, 1999.
3 V. Khripunov, Preliminary Nuclear Analysis of the IVVS penetrating the Outboard Shielding Blanket. NAG-
145-21-1-00. Garching, January, 2000.
4 A. Borisov, G. E. Shatalov, and S. Sheludjakov, Analysis of Neutron Streaming through the Viewing System
Straight Channel in the Divertor Cassettes of LAM RC/RTO Option. Nuclear Fusion Institute, RRC “Kurchatov
Institute”, Moscow, April-June 1999.
5 V. Khripunov, Neutronics Aspects of a Streaming through the Divertor Viewing Channel. Report at the
Remote Handling Interface Meeting, Garching, Garching, 25-30 June, 1999.
6 V. Khripunov, Radiation Fields in the Diagnostic System Environment. Report at the Safety/Remote
Handling Group Working Meeting for Occupational Safety. Garching JWS, 11 October, 2000.
7 G. E. Shatalov, N. N. Vasiliev, V. S. Zaveriaev, et al., Nuclear Analysis of the ITER Diagnostic Systems.
Contract work “Design of Specific Components". Report for the 3-d quarter, 2000. Nuclear Fusion Institute,
RRC “Kurchatov Institute”, Moscow, October 2000.
8 S. Sheludjakov and G. E. Shatalov, Diagnostic Channel in a Central Divertor Cassette. Report at the Progress
Meeting on ITER Nuclear Performance and Results for the Draft Nuclear Analysis Report. ITER Garching
JWS, 9-10 November, 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 6.5-1 Neutron Fluxes at the IVV Channel Entrance
Neutron Wall Loading ~ 0.4 MW/m 2
DT (14.1 MeV) 2.7 x 10 13
cm -2 s -1 ,
fast (En > 0.1 MeV) 9.7 x 10 13
cm -2 s -1 total (En > 0) ~1.8 x 10
,
14
cm -2 s -1 .
Neutron and gamma-ray fluxes were calculated at several locations along the channel axis
and at circular detectors ~0.5 m away from the axis (Figure 6.5-1).
The neutron flux attenuation along the 22-cm channel, that would allow direct viewing of the
interior of the plasma camera, is in the range of ~5 x10 -4 to 5 x 10 -3 at a distance of ~1.2 m
from the first wall to the rear divertor surface. The absolute flux levels at the edge of the
channel caused by the neutron leakage through the channel are significant, between 10 10 and
10 12 cm -2 s -1 :
Table 6.5-2 Neutron Fluxes at the End of the IVV Channel
DT (14.1 MeV) 1.4 x 10 10
fast (En > 0.1 MeV) 3.5 x 10 11
total (En > 0) ~9.2 x 10 11
cm -2 s -1 ,
cm -2 s -1 ,
cm -2 s -1 .
The penetration increases neutron fluxes inside and outside divertor port to a higher level (by
~10-20% locally). The neutron streaming effect spreads wide across all of the cryostat. The
peak fast and total flux values due to the effect of streaming through the channel are the next:
Table 6.5-3 Neutron Flux Peaks due to Streaming
through the IVV Channel, cm -2 s -1
n-fast n-total
at the inner port wall directly opposite
the unshielded duct (in a 1-m spot) 0.8-1.7 x 10 10
1.5-3.5 x10 10
in the vicinity of the lower port wall 4.5x10 8
7.9 x10 8
in the channel projecting to the side cryostat
surface
~7 x10 7
1.3 x10 8
The intense neutron fluxes lead to considerable activation of materials beyond the port walls,
and their correspondingly high dose rates at the cryostat are close to, or even exceed the
permissible dose rate limit after reactor shutdown.
High fast and total neutron flux fractions of ~2x10 7 and ~4.6 10 7 cm -2 s -1 , respectively, were
found at the cryostat near the divertor port door, due to this channel impact. It is ~5 % higher
the flux background in this region.
As a result, the residual dose rate at the cryostat flange directly opposite the divertor port (at
10 days after shutdown at the end of the DT-operation) is by ~50 µSv/hr higher than the local
background.
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ITER G 73 DDD 2 01-06-06 W0.1
6.5.2 Shielding Block Efficiency
A passively cooled gate type shielding enclosure, inserted in the IVV channel outside the port
volume, was proposed to reduce the streaming effects.
As shown in Figure 6.5-2, it is locked by a steel “shutter” in depth 100 mm in the cannel and
by an additional 150 mm block from outside.
Figure 6.5-2 Steel Shielding Block closing the IVV Channel
The performance of this shutter to achieve lower neutron flux level at the channel exit was
investigated in reference 1 . The radial and axial distributions of neutron and photon fluxes,
and nuclear heat deposition were calculated in shutter layers (as shown on the right-hand side
in Figure 6.5-2) and outside the divertor (Table 6.5-4).
Table 6.5-4 Fluxes and Energy Depositions in the Steel Shutter
Front
Rear
(surface) layer 3-cm layer
R < 11 cm *)
(in the channel)
R < 11 cm R ~ 0.5 m Units
DT-neutron flux 3.6 x 10 10
< 1 x 10 9
- 4 x 10 9
cm -2 s -1
Fast neutron flux 1.7 x 10 12
2.1 x 10 11
- 7.9 x 10 11
cm -2 s -1
Total neutron flux 5.1 x 10 12
0.4 x 10 12
- 1.3 x 10 12
cm -2 s -1
Total photon flux ~ 2.5 x 10 12
0.5 x 10 11
- 2.9 x 10 11
cm -2 s -1
Nuclear energy
deposition in steel
~ 0.13 0.002 - 0.017 W/cm 3
*) R is a distance from the channel axis.
1 G.E. Shatalov, A.A. Borisov, I.A. Kartashev, A.G. Serikov, S.V. Sheludyakov, O.L. Schipakin, Neutronic
Analysis of the ITER Vacuum Vessel/Cryostat Environment. Report of RF HT for the 4 th Quarter 2000, JF-04-
00/4. September, December 2000.
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ITER G 73 DDD 2 01-06-06 W0.1
The volumetric nuclear energy deposition range along the shutter from 0.13 W/cm 3 to ~2
mW/cm 3 . At the last rear layer it increases in ~10 times in radial direction from the channel
axis (R < 11 cm) to the blanket (R ~ 0.5 m).
The integrated nuclear heating is ~ 580 W from which ~550 W is the photon energy and the
rest 30 W is the neutron energy.
Estimated plasma heat radiation at the shutter front surface is ~0.3 W/cm 2 . Thus the total
power released in this shielding structure is ~ 0.7 kW, and a passive cooling might be
possible.
Additional shielding effect may be achieved by adding water in its composition. As an
option, a 250-mm 60%SS/40%H2O shutter was considered. In this case, the neutron flux
attenuation along the cannel is ~2 times higher than in the case of a pure steel. However, due
to increased neutron absorption and secondary photon production the total nuclear heating in
this closure is ~240 W higher.
6.5.3 Findings
Accurate 3D-modeling and calculations showed that an expected shielding efficiency of a
closure, i.e., a reduction of streaming effects by more than one order of magnitude, can not be
realized in the current design. Neutrons entering in the channel from the plasma region
stream round a channel closure.
As was mentioned, the effects of the viewing channel in a divertor increases radiation
environment in and around the port by ~ 20% (and at the cryostat door ~ 5%). At the same
time, neutron streaming through the gaps between the divertor cassettes, a wedge type gap
below the blanket modules 17, and through the pumping slot dominates even in the absence
of the IVV channel.
The IVV integration at the divertor level is on going and further design modification should
be considered in a frame of a more general problem of the divertor port shielding to diminish
the radiation condition in the port surrounding by about one order of magnitude.
6.6 Photo-Neutrons in the Plasma Chamber
Neutron and photon fields in the plasma chamber and surrounding structures were
investigated in toroidal geometry using the MCNP 4B code system 1 and a multilayered
model of the first wall (a 1 cm Be-coverage bonded to a 2 cm copper layer including water
channels, then to a 2 cm steel structure) followed by the steel/water shielding blanket and the
vacuum vessel (See references 2 and 3 ).
1 J. F. Briesmeister, J. S. Hendricks, H. M. Abhold, J. D. Court, S. Frankle, B. L. Kirk, “MCNP 4B, Monte
Carlo N-Particle Transport Code System,” LANL, Los Alamos, New Mexico. LA-12625-M, RSIC CCC-660
(1997).
2 V. Khripunov, The First Wall “ Sandwich” as a Source of Photo-Neutrons in ITER. IDoMS No.: NAG-143-
11-1-00. Garching, 13 January, 2000.
3 V. Khripunov, The ITER First Wall as a Source of Photo-Neutrons. Report submitted to the 21 st Symposium
on Fusion Technology, (SOFT-21), September 11-15, 2000, Madrid, Spain. To be published in Fusion
Engineering and Design.
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ITER G 73 DDD 2 01-06-06 W0.1
These investigations, in particular, show that some products of nuclear reactions in the copper
and steel layers of the first wall, such as 9 Be, 56 Fe, 63 Cu, 65 Cu, and 59 Ni, emit gamma-rays
with energies higher than 1.665 MeV - the threshold for the 9 Be (γ, n) 8 Be (= 2 4 He) photoneutron
reaction in beryllium. As a result of this reaction, the fusion neutron production in the
plasma chamber, will be accompanied by the prompt photo-neutron yield in the Be/Cu/SSfirst
wall “sandwich”.
A non-zero source of delayed photo-neutrons born by the residual photons will be present in
the reactor core also during dwell times and after shutdown.
6.6.1 Incident neutrons and secondary photons
The following values of the incident 14.1 MeV neutron yield are typical within the plasma
chamber during D-T-plasma burning at the nominal fusion power of ~ 500 MW:
the specific neutron source strength ~2.5 10 11 n (14.1 MeV)/cm 3 /s;
the total neutron source in the plasma volume ~1.8 10 20 n (14.1 MeV)/s.
The neutron and secondary photon fluxes and consequently the expected prompt photoneutron
production rate are maximal in the first wall region and are proportional to the fusion
power (Table 6.6-1).
Table 6.6-1 Incident Neutron and Photon Fluxes in the Plasma Chamber
during the Nominal D-T Operation at the Fusion Power ~500 MW
Energy, MeV Neutron Fluxes, cm -2 s -
Photon Fluxes, cm -2 s -1
14.1 3.3 10 13
> 2 6.2 10 13
> 1.66 6.5 10 13
> 1 7.7 10 13
> 0.1 1.3 10 14
> 0 (total) 2.0 10 14
About 24 % photons may produce prompt photo-neutrons in Beryllium.
Nuclear Analysis Report Page 157
1
2.6 10 13
1.1 10 14
It should be noted also, that an energetic tail, comprising ~10 9 cm -2 s -1 (E γ > 15 MeV) of
which 10% exceeds 22 MeV, is appeared in the photon flux spectrum as a result of the high
energy primary neutron captures (En>10 MeV) accompanied by strong radionuclide
transitions (~8 MeV or more).
6.6.2 Prompt photo-neutrons
Using the gamma-ray spectrum and the Be (γ, n)-reaction cross sections, which is of the order
of a millibarn in a wide gamma-energy range, a prompt photo-neutron production rate of ~3.1
10 9 photo-neutrons/cm 3 s was calculated in the Be- plasma facing layer as a result of energetic
photons from the surrounding structural materials. The total prompt photo-neutron yield is ~
2 10 16 n/s.

ITER G 73 DDD 2 01-06-06 W0.1
The energy distribution of the prompt photo-neutron source in the Be-layer is given in Figure
6.6-1 and the prompt photo-neutron flux spectrum in the plasma chamber is compared with
the direct neutron flux spectrum in Figure 6.6-2.
1x10 9
1x10 8
1x10 7
1x10 6
1x10 5
1x10 4
1x10 3
1x10 2
1x10 1
1x10 0
0 2 4 6 8 10 12 14 16 18 20
Neutron Energy, MeV
Figure 6.6-1 The prompt photo-neutron
source in the Be-first wall
Nuclear Analysis Report Page 158
1x10 16
1x10 15
1x10 14
1x10 13
1x10 12
1x10 11
1x10 10
1x10 9
1x10 8
1x10 7
1x10 6
1x10 5
1x10 4
DT-Neutron Source in the Plasma Chamber
Photo-Neutron Source in the Be-First Wall
1x10 -1 1x10 0 1x10 1 1x10 2 1x10 3 1x10 4 1x10 5 1x10 6 1x10 7 1x10 8
Neutron Energy, eV
Figure 6.6-2 Neutron flux spectra at the FW
from the incident (D-T) plasma source
and the prompt photo-neutron source
The intensity of the prompt photo-neutron emission (and derived secondary photon emission)
as well as corresponding neutron fluxes and nuclear responses in the first wall (such as
nuclear energy deposition, material damage, gas-production He-production in structural
materials and absorbed dose rates) are four orders of magnitude lower than for the incident
D-T neutron source under normal operating conditions.
The neutron shielding characteristics built into the primary bulk shield attenuate the fusion
neutrons sufficiently, by ~7 - 8 orders of magnitude from the first wall to the space inside the
cryostat.
Thus the impact of photo-neutrons on the nuclear performance as a whole is practically
negligible.
6.6.3 Impact on neutron diagnostics
Neutron spectrometry will be a key element of the plasma diagnostic system in ITER 1 . The
diagnostic information will be derived from the direct neutron emission and from the
observation of small deviations from the “normal” shape of the thermonuclear peaks (a
Doppler width) and their shifts.
The observed neutron energy distribution will represent spectra superposition reflecting
different peculiarities of reactions in plasma including minority contributions from the D-D
and T-T reactions (~10 -2 and ~10 -4 ) over the regions 3.5 - 21.5 MeV.
1 V. Mukhovatov, A.E. Costley et al., ITER physics program and implications for plasma measurements.
Rev. Sci. Instrum. 68 (2) (1997) pp 1250-1255.

ITER G 73 DDD 2 01-06-06 W0.1
Presently envisioned measurements of the plasma temperature and the D-T reaction rate
profiles, using neutron cameras, cover the plasma periphery where there is low neutron
emission typically 2-3 orders of magnitude lower than average values. In this area the
scattered background superimposed on the main thermonuclear neutron energy distribution is
important for neutron spectroscopy 1 .
A consideration shows that photo-neutron wall emission is expected to play important role in
the diagnosis and control of thermonuclear plasmas in ITER.
Most of the prompt photo-neutrons are fast neutrons (Ephoto-n > 0.1 MeV). The high energy
tail of the photon spectrum will probably generate “supra” photo- neutrons of a very high
energy. The estimated production rate of such “supra” photo-neutrons in 1 cm 3 Be in the first
wall is ~ 10 5 cm -3 s -1 (Ephoto-n > 14 MeV) and ~ 10 4 cm -3 s -1 (Ephoto-n ~ 18 MeV) (Figure 6.6-1).
The total prompt photo-neutron flux expected in the plasma chamber during plasma burning
is ~3.2 10 10 cm -2 s -1 . The estimated “supra” energetic neutron flux is ~ 10 5 cm -2 s -1 . These may
be not negligible from the diagnostic viewpoint.
Thus the estimated photo-neutron wall emission in this energy range in addition to the
scattered background 1 accompanying the observation of fusion plasma neutrons becomes an
essential factor in high accuracy neutron diagnostics.
The delayed photo-neutron background (See below) may be also important for calibration
issues.
6.6.4 Delayed photo-neutrons
Unstable nuclei appearing in the first wall during irradiation, such as 62 Cu, 64 Cu, 54 Mn, 58 Co,
and 60 Co, will also emit high energy gammas both during dwell time and after shutdown. The
maximum residual gamma-source intensity ~3 10 12 cm -3 s -1 will be achieved at the end of the
DT-operational period (~0.3 MWa/m 2 ). In a one week - one month cooling period this
decreases from ~ 3 10 12 cm -3 s -1 to ~2 10 11 cm -3 s -1 . See Figure 6.6-3.
A part of the decayed photons exceeding the Be (γ, n)-reaction threshold varies from ~40 %
during the first hour after shutdown to 0.1 % during a 7-30 days maintenance period and
further to ~ 0.01 % during one cooling year (Figure 6.6-4).
1 P. Antozzi, G. Gorini, J. Källne, N. Olsson, E. Ramström, M. Campanella, Scattering effects in neutron
diagnosis of DT tokamak plasmas. Rev. Sci. Instrum. 66 (1) (1995) 939-941.
Nuclear Analysis Report Page 159

ITER G 73 DDD 2 01-06-06 W0.1
1x10 14
1x10 13
1x10 12
1x10 11
1x10 10
1x10 9
1x10 8
1x10 7
1x10 6
1x10 5
1x10 4
1x10 3
1x10 2
1x10 1
1x10 0
Cu-62, Cu-64 Co-58, Mn-54 Co-60 Nb-94
E > 1.66 MeV
total
1 day 1 yr 10 yr100 yr
Cooling Time after Shutdown, s
Figure 6.6-3 Residual _-ray production
in the FW as a function of cooling time
after the DT-phase (0.3 MWa/m 2 )
Nuclear Analysis Report Page 160
1x10 12
1x10 11
1x10 10
1x10 9
1x10 8
1x10 7
1x10 6
1x10 5
1x10 4
1x10 3
1x10 2
1x10 1
1x10 0
14 days
1 year
Threshold
Energy
1 hr
1 day
1 min
1 s
0 2 4 6 8 10 12 14
Residual Gamma-Ray Energy, MeV
Figure 6.6-4 Residual _-ray spectra
after shut down
The “delayed” photo-neutrons produced by those residual photons in the Be-first wall
coverage will appear in the plasma chamber in the meantime.
The yield of the delayed photo-neutrons depends on the intensity of the residual gammasources
accumulated in the irradiated first wall, and is a function of the preceding
accumulated neutron fluence, operation scenario, and cooling time. Thus the delayed photoneutron
production rate will increase during operation and decrease during dwell time and
after shut-down following the variation of the residual gamma-source.
The maximum delayed photo-neutron source intensity is ~ 8 10 6 cm -3 s -1 at the end of the D-T
operation period. Within one year it decreases by about 5 orders of magnitude, by the same
ratio as the gammas, which causes photo-neutron production to decrease.
As a result, a maximum (non-negligible) delayed photo-neutron flux ~10 8 cm -2 s -1 may be
expected in the plasma chamber in the dwell phase ~1000 s after shutdown. It decreases to ~
10 5 cm -2 s -1 after one cooling year.
The delay neutron flux and spectrum variations as functions of cooling time at the end of the
D-T phase are shown in Figures 6.6-5 and 6.6-6.

ITER G 73 DDD 2 01-06-06 W0.1
1x10 8
1x10 7
1x10 6
1x10 5
1x10 4
1x10 3
1x10 2
1x10 1
1x10 0
1x10 -1
1x10 -2
n-total
n-fast
gamma-total
n-thermal
1 day 1 yr 10 yr 100 yr
1x10 0 1x10 1 1x10 2 1x10 3 1x10 4 1x10 5 1x10 6 1x10 7 1x10 8 1x10 9 1x10 10
Cooling Time after Shutdown, s
Figure 6.6-5 Delayed photo neutron
fluxes and total secondary photon fluxes
as a function of cooling time
Nuclear Analysis Report Page 161
1x10 8
1x10 7
1x10 6
1x10 5
1x10 4
1x10 3
1x10 2
1x10 1
1x10 0
1x10 -1
6.6.5 Effects of the delayed photo-neutrons
10 years
1 year
1x10 -1 1x10 0 1x10 1 1x10 2 1x10 3 1x10 4 1x10 5 1x10 6 1x10 7
Neutron Energy, eV
1 s
14 days
Figure 6.6-6 Evolution of the delayed
photo-neutron spectrum after shutdown
The delayed photo-neutron fluxes and the corresponding nuclear responses are 7-9 orders of
magnitude lower than caused by the incident 14.1-MeV-neutrons that is typical for regions
behind the bulk radiation shield and the biological shield.
In intermediate periods of the D-T-campaign they will be several times lower.
The maximum additional neutron fluence from the delayed neutrons is ~1.7 10 13 cm -2 . 40 %
of that is the fast (> 0.1 MeV) component. Half of the delayed neutron fluence will be
reached during the first 6 months after reactor shutdown, and the rest - in 100 years.
The estimated dose rate in neutron sensitive detectors and steel due to the delayed photoneutron
absorption is about ~3 and ~0.6 Gy/hr, respectively. These values decrease by 2-3
orders of magnitude within 2 weeks.
Postponed activation of “pure”, non-irradiated materials by the delayed photo-neutrons is in
principle possible. Such materials, as instruments and equipment for dismounting and repair,
for detectors for diagnostics, as well as the coolant and atmosphere, will probably be in
contact with the first wall during maintenance periods. As an example, the dose rate ~ 15
µSv/hr was evaluated at a steel container irradiated by the residual photo-neutrons emitted
from a blanket module for 3 months. It decreases to ~0.1–0.4 µSv/hr after one day, and to
0.01-0.07 µSv/hr one year later.
Based on this consideration, it may be concluded that the delayed photo-neutrons will not be
a problem outside the machine during dwell time longer than 1000 s. However, they should
be taken into account for procedures inside the chamber, and when open channels looking at
the first wall surface are used.

ITER G 73 DDD 2 01-06-06 W0.1
7 Dose rate estimate outside bioshield
Dose rate estimation outside the bioshield is necessary from the view points of occupational
safety and structural design of reactor building which provides radiation shielding not only
for the workers but also for public. The estimation may be required for the three different
plant states; namely, during reactor shutdown, during transportation of radioactive
components for their replacement and during reactor on power. In the estimation for the first
and third plant status, radiation streaming through the penetrations in the bioshield can play
dominating role.
There is no design limit on the dose rate in the space between the bioshield and the outer
wall of the reactor building during operation, since no personnel access in the reactor
building is expected. On the other hand, during shutdown, dose rate outside the bioshield
should be less than 10 μSv/h from the occupational safety point of view.
Outside the reactor building, dose rate value should be always less than 0.5 μSv/h
independently from the plant state. In the actual building design, shielding optimisation will
be required not only satisfying the above design limits but also giving reasonably low cost of
the reactor building.
7.1 Dose rate during reactor shutdown
The bioshield thickness is generally determined by structural reasons and not by shielding
requirements. As stated in the section 3.6, shielding requirement for the bioshield depends on
how soon personnel access is expected. It is sufficient to provide an attenuation of a factor of
10 so as to reduce the dose rate to 10 μSv/h 10 6 s (~ 12 days) after shutdown assuming that
the required 100 μSv/h is achieved in the cryostat. For the place where more quick access
(less than one day after shutdown) is required, a factor of 100 may be necessary, since dose
rate at the shutdown is higher by about an order of magnitude than that 10 6 s after shutdown.
The actual bioshield has a thickness of 2 m and provide ~ 8 orders of magnitude attenuation
if no penetration exists in the bioshield.
Many penetrations are necessary to route hydraulic, electrical guides, etc. in the bioshield.
Preliminary estimation shows (see Figure 7.1-1) that a hole with 50 cm radius provides an
order of magnitude attenuation and that with 15 cm two orders of magnitudes. Most of the
holes in the bio-shield have a radius less than 15 cm, for example the IC H & CD
transmission line (R < 10 cm) and the EC H & CD wave guides (R < 10 cm).
In some cases, the size of penetration exceeds the above 15 cm or even 50 cm radius but
filling inside with shielding material is possible. As a typical example of such cases,
penetrations for blanket cooling pipes have a 64 cm radius, including 36 cooling pipes.
Proper shielding material should be provided among pipes in the hole in order to give
required attenuation. Figure.7.1-2 shows how much thickness is required for such shielding
material filled in the hole, depending on the diameter of cooling pipes.
The following penetrations requires careful shielding design (or maintenance plan to limit the
personnel access) since filling in the penetration with shielding material may not be possible
differing from that for blanket cooling pipes.
Nuclear Analysis Report Page 162

ITER G 73 DDD 2 01-06-06 W0.1
• electric current and cryogen feed through for the magnet (85 cm diameter; 30)
• He cooling gas pipes for the thermal shield (100 cm diameter; 2)
• water drain pipes for vacuum vessel (32 cm diameter; 2)
If access to the space where these pipes locate (the pipe chase area) is required, shutdown
dose rate analysis is required in future. At present, only dose rate estimate during operation
has been conducted (see section 7.2).
Relative dose attenuation at the end of
hole
1.000
0.100
0.010
0.001
0 10 20 30 40 50 60 70
Hole radius R (cm)
Figure 7.1-1 Relative attenuation of
dose rate through the hole with radius
R (L=200 cm) (by the Handy method)
Note: Hole should not see the collimated radiation flux
Relative attenuation
0.01
0 10 20 30 40
Nuclear Analysis Report Page 163
1.00
0.10
R=2cm
R=4cm
R=6cm
Distance from the hole entrance (cm)
Figure 7.1-2 Relative dose attenuation
through hole with radius R when SS
shield is filled between pipes.(by the
Handy Method)
7.2 Dose rate during maintenance condition
As a typical case of maintenance condition, blanket module replacement procedure has been
studied. A 3D model was created, in which two failed blanket modules are contained in a
cask placed inside the maintenance pit. In order to make personnel occupation possible on the
floors above and below the equatorial level independently from maintenance work in the
equatorial pit, the dose rate, contributed from radiation source in the pit, should be < 10
μSv/h on those floors.

ITER G 73 DDD 2 01-06-06 W0.1
7.2.1 Model description
A A
B
B
Sec. B-B
Transfer
Cask
Sec. A-A
Figure 7.2-1 Vertical (left & B-B) and horizontal (A-A) sections of the equatorial
maintenance pit
A 3-D model for MCNP code has been created based on the drawings provide by the remote
handling group in Naka JWS on 12 May 2000 1 . In Figures 7.2-1 the vertical sections of the
maintenance pit are shown. Dose rates are evaluated being averaged in the cyan spherical
cells in this figure. A 3-D view of the model is shown in Figure.7.2-2 which is obtained with
the SABRINA code. The blanket modules have the dimension of 119 cm / 45.1 cm / 123 cm
(blanket module #14 on the outer equatorial plane).
The cask wall thickness is assumed to be 6 cm in this study. The correct thickness is 1.7 cm
and will give higher dose rate by a factor of ~5 (an order of attenuation by 6 cm stainless
steel).
1 Toshiyuki Suzuki, Remote Handling Group, Naka JWS, Private communication (by mail),
12 May 2000.
Nuclear Analysis Report Page 164
y
Blanket
Module
x

ITER G 73 DDD 2 01-06-06 W0.1
A decay gamma-ray source distribution at 10 6 s after M-DRG1 operation (0.3 MWa/m 2 ) is
produced using an Sn (ANISN) transport and an activation (ACT-4) codes with 1-D annulus
model of the reactor. The spatial distribution in each layers (48 layers in radial direction) of
the outboard blanket module and the energy spectrum (54 energy groups between 0.03 to 3
MeV) of the decay gamma-ray was taken into account. The two blanket modules are placed
in the transfer cask with the first wall upwards.
Figure 7.2-2 3-D view of the modules inside the maintenance pit.
7.2.2 Results and discussion
Blanket
Modules
The dose rates in the rooms above and below the equatorial maintenance pit are shown in
Figure 7.2-3. The dose rate level is low enough to satisfy the limit (10 μSv/h) in the room
below the pit. However, in the room above the pit, it is around the limiting value.
When we take account of the fact that the real cask has a thinner wall by 4.3 cm, actual dose
rate can be about 50 μSv/h. It might be required to have some shielding reducing dose rate
about one order of magnitudes to fully access in the room above the pit when two activated
blanket modules are extracted from the torus into the pit.
Figure 7.2-3 (and Figure 7.2-4) also shows dose rates inside and outside the transfer cask.
Dose rate inside the cask but not just above the blanket module is around 10 Sv/h, which is of
coarse too high for personnel to access. The maximum dose rate outside the cask should be
around ~1 Sv/h, when we make correction concerning cask wall thickness (1.317x 5 = 7
Sv/h).The dose rate in the building can be ~10 Sv/h, when a cask containing these activated
blanket modules move in the reactor building. It should be attenuated by about 7 orders of
magnitude by the reactor building (~ 1.4m thick) since the dose rate outside the building
should be less than 0.5 μSv/h.
Nuclear Analysis Report Page 165

ITER G 73 DDD 2 01-06-06 W0.1
0.1 2.5 8.2 10.8 7.0 _ μSv/h
10.6
Sv/h
9.5
0.003 0.1 0.4 0.4 0.2 _ μSv/h
Figure 7.2-3 Dose rates inside, outside the cask and in the room above and below
the equatorial floor.
50 57 63 68 71 74 75 75 74 72 69 64 59 53Sv/h
Blanket
Module
mSv/h
Figure 7.2-4 Dose rate just above the activated blanket modules
Nuclear Analysis Report Page 166
234
301
308
430
612
606
846
1052
1128
1317

ITER G 73 DDD 2 01-06-06 W0.1
7.3 Dose rate outside the bioshield during operation
7.3.1 Calculation model
A 3-D analysis has been conducted for estimating the effect of penetrations in the bioshield
on the dose rate during reactor operation. Dimensions of major penetrations (32, 85 and 100
cm diameters; see 7.1) were given by the Nuclear Technology Division in Naka JCT. A 3-D
geometry input for MCNP has been created simulating the geometry of the pipe chase area of
the reactor building as shown in Figures 7.3-1(a) through 7.3-1(c). Building dimensions are
decided based on the drawings of ‘FEAT10006800012D0006W’ and ‘62.0333.000 1~
11_.2D.02.00R’.
Neutron source of 175 energy groups (current J+), which was obtained by the 1-D
calculation, has been placed on the inner surface of the cryostat. Gamma source current at
inner surface of the cryostat was neglected since its contribution was confirmed to be
insignificant.
Cryostat
Absober
Bioshield
Penetration
Figure 7.3-1 (a) 3-D view of the pipe chase area model. Pictured by SABRINA
A A
Figure 7.3-1 (b) Vertical cross section of the 3-D model
Outer Wall of the Building
Nuclear Analysis Report Page 167

ITER G 73 DDD 2 01-06-06 W0.1
Cryostat Bioshield
Penetration
Figure 7.3-1 (c) Horizontal cross section of the 3-D model (A-A)
7.3.2 Calculation results and discussion
The following 3 cases of calculation have been conducted changing penetration diameter.
• Case 1 : 32 cm of diameter (water drain pipes for vacuum vessel)
• Case 2 : 85 cm of diameter (electric current and cryogen feed through for the magnet)
• Case 3 : 100 cm of diameter(He cooling gas pipes for the thermal shield)
1.E+08
1.E+07
1.E+06
1.E+05
1.E+04
1.E+03
1.E+02
1.E+01
1.E+00
Reflecting
Boundary
Cryostat
1-D
3-D(32 cm)
3-D(85 cm)
3-D(100 cm)
Outer Wall of the Building
Bioshield
1.E-01
-50 0 50 100 150 200
Distance from Bioshield Inner Surface (cm)
Figure 7.3-2 Dose attenuation along the penetration axis in the bioshield
Nuclear Analysis Report Page 168

ITER G 73 DDD 2 01-06-06 W0.1
Dose rate attenuation along the penetration axis are shown in Figure 7.3-2. When we do not
have any penetration, ~ 7 orders of magnitude attenuation can be expected by the bioshield.
When a penetration exists, much less dose attenuation is obtained. 100 cm of diameter gives
about an order of attenuation and 32 cm two orders of attenuation. These results are very
consistent with those predicted by the Handy Method (see 7.1).
Dose rates in the pipe chase area for the above three cases are shown in the Figure 7.3-3(a),
(b) and (c). Depending on the penetration sizes, dose rates at the inner surface of the building
wall are in the range of 1.3x10 -4 ~ 2.2x10 -3 Sv/h. In order to satisfy the limit of 0.5 μ Sv/h at
the outside of the building, four orders of magnitude of attenuation is required for the
building wall. When no penetration exists in the building wall, this attenuation is obtained by
~1.2 m thick concrete.
Figure 7.3-4 shows the horizontal distribution of dose rate in the building for the case 3 (100
cm). As we can see in this figure, the dose distribution in the building is fairly flat.
8.30-8
6.09-4
Figure 7.3-3 (a) Dose rate distribution in vertical cross section for the case 1 (32cm)
6.96-6
5.55-8
4.44-4
5.00-6
2.94-2 1.24-2
1.03-5
2.33-4
3.38-4
3.32-3
2.21-5
1.27-4
5.11-4
1.49-3
Sv/h
Sv/h
Figure 7.3-3 (b) Dose rate distribution in vertical cross section for the case 2 (85cm)
Nuclear Analysis Report Page 169

ITER G 73 DDD 2 01-06-06 W0.1
1.81-5
8.73-6
5.66-4
8.28-4
5.21-2 1.95-2 4.98-3
2.19-3
Figure 7.3-3 (c) Dose rate distribution in vertical cross section for the case 3 (100cm)
3.65-3
3.48-3
9.66-3
8.52-3
5.21-2 1.95-2
8.43-3
9.46-3
3.47-3
3.56-3
2.09-3
3.65-3
4.98-3
3.70-3
Sv/h
2.02-3
Sv/h
1.42-3
1.80-3
2.19-3
1.85-3
1.47-3
Figure 7.3-4 (a) Dose rate distribution in horizontal plane (z=0: penetration axis) for the
case 3 (100cm)
Nuclear Analysis Report Page 170

ITER G 73 DDD 2 01-06-06 W0.1
8.38-6
1.81-5
1.21-5
6.47-6
8.73-6
8.27-6
Sv/h
5.22-4
5.66-4
5.03-4
7.12-4
8.28-4
6.89-4
Figure 7.3-4 (b) Dose rate distribution in horizontal plane (z=325) for the case 3 (100cm)
Nuclear Analysis Report Page 171

ITER G 73 DDD 2 01-06-06 W0.1
8 The DD Phase Nuclear Performance
Deuterium plasma will be used during an early phase of ITER operation to test physics
performance, to prepare of reliable prototype scenarios for a subsequent D-T operation, to
commissioning fusion diagnostic and other equipment. Half a year or more will likely be
allocated to long (~300-s) pulses in advanced D-D H-mode plasmas with sufficient auxiliary
heating (~100 MW).
8.1 Two component neutron source
The analysis 1 shows that in advanced D-D H-mode regimes, proceeding the DT phase, the
2.45-MeV neutron yield ~ 3.5 x 10 18 n/s from the reaction
D+D (50%) → nDD (2.45 MeV) + 3 He (0.82 MeV)
will be accompanied by considerable tritium production from the second branch of the D-Dreaction
D+D (50%) → T (1.01 MeV) + 1 H (3.03 MeV).
The total tritium production during ~ 3 10 5 s of the D-D-burning (300 s x 10 3 pulses) is about
5.2 g. Most of it, however, burns immediately.
Assuming conservatively that all the generated tritium in the equilibrium plasma at the
highest auxiliary heating power ~100 MW and with a good particle confinement will react at
high temperature ~9 keV with the available deuterium, then about 3.5 x 10 18 DT (14.1 MeV)
neutrons per second will be produced in the plasma chamber 1 .
In the opposite case of a bad confinement and maximum gas removal (pumping) efficiency,
estimates 2 indicate a five times lower (~7 10 17 n/s) 14.1-MeV neutron yield.
But in all cases this two component (2.45- and 14.1 MeV) neutron source will influent the
ITER nuclear performance causing significant nuclear heat deposition in structures, and
activation of plasma facing structures, leading to conditions which cannot be ignored from
the radiation safety and maintenance standpoints 2 3 .
8.2 Neutron spectra and flux distributions
Multigroup (175n+42γ) discrete ordinate calculations were performed to determine neutron
and photon fluxes and energy distributions in the steel/water shielding blanket and vacuum
1 D. Boucher, Neutron /Tritium Rates During D-D Phase in RTORC-ITER. Interoffice Memorandum RCI_169,
San-Diego JWS, 6 November 1998.
2 V. Khripunov, The DD-Phase Activation of the LAM. G 73 RI 101 98-11-30 W 0.1 (NAG-121-25-11-98).
ITER Garching JWS, November, 1998.
3 V. Khripunov, Nuclear Performance of the D-D Phase of ITER. Proceedings of the Fifth International
Symposium on Fusion Nuclear Technology, Rome, Italy, 19-24 September, 1999, Part B. Fusion Engineering
and Design 51-52 (2000) 281-287.
Nuclear Analysis Report Page 172

ITER G 73 DDD 2 01-06-06 W0.1
vessel from the combined neutron source. The DANTSYS code system 1 , the P3-S8
approximation, FENDL-2 cross section library 2 and a simple one-dimensional “toroidal”
cylinder model of the reactor were used for that.
Neutron spectra shown in Figure 8-1 are similar for both the nominal D-T operations and
preceding D-D operations. However, an additional 2.45-MeV peak is present in the D-D case.
As well as in the D-T phase spectrum, the fast neutron fraction in the D-D spectrum is ~ 60%
(Table 8-1). More than a half of the total neutron flux (in case of the minimum pumping
efficiency, i.e. maximum tritium burn-up) is caused by the D-T-neutron source component.
1x10 15
1x10 14
1x10 13
1x10 12
1x10 11
1x10 10
1x10 9
1x10 8
1x10 7
1x10 6
1x10 5
n (DT-phase)
n (DD-phase)
γ (DT-phase)
γ (DD-phase)
1x10 -1 1x10 0 1x10 1 1x10 2 1x10 3 1x10 4 1x10 5 1x10 6 1x10 7 1x10 8
Energy, eV
Figure 8-1 Neutron and Photon Spectra at the First Wall
Table 8-1 Fast and Total Fluxes at the First Wall
for the Advanced D-D and the Nominal D-T Operation Regimes, cm -2 s -1
Phase D-D D-T
Source DD-n DT-n DD & DT DT-n
n-14.1 - 1.1-3.3 10 11
1.1-3.3 10 11
~ 3.4 10 13
n-2.45 2.4 10 11
2.7-8.1 10 9
2.4-2.5 10 11
~ 9 10 11
n-fast (>0.1) 1.4 10 12
0.5-1.7 10 12
1.9-3.1 10 12
~1.3 10 14
n-total 2.3 10 12
0.7-2.9 10 12
3.0-5.2 10 12
~2 10 14
γ-total 8.6 10 11
0.4-1.8 10 12
1.3-2.7 10 12
~ 1 10 14
The radial neutron flux distributions in the outboard bulk radiation shield are shown in Figure
8-2. The distributions produced by the 2.45 MeV neutrons are presented separately.
1 R. E. Alcouffe, R. S. Baker et al., DANTSYS 3.0 One-, Two-, and Three-Dimensional, Multigroup, Discrete
Ordinates Transport Code System, LANL, Los Alamos. RSIC Computer code collection CCC-547, August
1995.
2 A. Pashshenko and H. Wienke, FENDL/E-2.0, Evaluated Nuclear Data Library of Neutron Nuclear
Interaction Cross Sections and Photon Production Cross Sections and Photon-Atom Interaction Cross Sections
for Fusion Applications, Report IAEA-NDS-175, International Atomic Energy Agency, March 1997.
Nuclear Analysis Report Page 173

ITER G 73 DDD 2 01-06-06 W0.1
1x10 13
First Wall Shielding Blanket Gap Vacuum Vessel
1x10 12
1x10 11
1x10 10
1x10 9
1x10 8
1x10 7
1x10 6
1x10 5
1x10 4
n(DD+DT)-total
n(DD+DT)-fast
n(DD)-total
n(DD)-fast
900 910 920 930 940 950 960 970 980 990 1000
Radius, cm
Figure 8-2 Neutron Flux Distributions in the Outboard Radiation Shield
for the D-D and D-T Neutron Source Components (D-D phase)
It is clear from these curves that the resultant bulk shielding efficiency is determined by the
14.1-MeV source neutrons: the neutron flux levels behind the blanket and vacuum vessel
are, correspondingly, one and two order of magnitude higher than in the case of neglecting
accompanying DT-reactions.
8.3 Neutron wall loading and fluence
Nuclear power, neutron wall loading and other key values are compared in Tables 8-2,-3,-4
for both the advanced D-D regime and the nominal D-T-regime. Upper values in the D-D
case correspond to the maximum tritium burn-up rate and good particle confinement.
It is seen from Table 8-2 that the 14.1-MeV source neutrons increase the neutron wall loading
and neutron fluence by up to a factor of 6 in comparison with the 2.45 MeV neutrons.
Table 8-2 Neutron Yield, Wall Loading and Power
for two Operation Regimes
Source Energy, D-D D-T Units
Particles MeV (H-mode) (Nominal)
Fusion Power n+charged 17.6 - 500 MW
particles 11.3-27.1 6.3-15 - MW
Neutron Power DD-n 2.45 ~1.4 - MW
DT-n 14.1 ~1.6 - 8 ~400 MW
Neutron Yield DD-n 2.45 3.5 10 18
- s -1
DT-n 14.1 0.7-3.5 10 18
1.8 10 20
s -1
Neutron Current DD-n 2.45 4 10 11
- cm -2 s -1
DT-n 14.1 0.8-4 10 11
~2.4 10 13
cm -2 s -1
Neutron Wall Loading DD-n 2.45 0.0015 - MW/m 2
DT-n 14.1 0.002-0.009 ~0.55 MW/m 2
Neutron Fluence*) DD-n 2.45 1.4 10 -5
- MWa/m 2
DT-n 14.1 ~ 2-9 10 -5
~0.3 MWa/m 2
*) ~300 s x 10 3 pulses of DD-plasma burning are assumed here.
Nuclear Analysis Report Page 174

ITER G 73 DDD 2 01-06-06 W0.1
The main nuclear characteristics experienced in the proposed D-D operation phase, may be as
high as: the fast (> 0.1 MeV) neutron flux of ~ 2-4 x 10 12 n/cm 2 s, the neutron wall loading ~
0.003-0.01 MW/m 2 , and the first wall neutron fluence ~ 10 -5 MWa/m 2 (or

ITER G 73 DDD 2 01-06-06 W0.1
8.5 D-D phase activation
Radiation conditions inside the plasma chamber and outside the vacuum vessel were
evaluated by successively using DANTSYS 1 and EASY 1 code systems for different cooling
periods as a function of D-D pulse number (ten 300-s pulses per day).
As was established earlier 2 , the residual dose rates in the plasma chamber are determined
mainly by the first wall activation. They differ by a factor 2 or 3 for the minimum and
maximum tritium burn-up cases examined (Figure 8-3).
10
1
0.1
0.01
0.001
0.0001
0.00001
10 000
1 10 100 1000
Cooling Time, days after shutdown
Nuclear Analysis Report Page 176
1000
100
10
1 pulse
(300 sec)
Figure 8-3 Dose Rates in the Plasma Chamber
as a Function of the D-D Pulse Number
(each band corresponds to the maximum and minimum tritium burn-up)
The main contributor to the dose rate 1-3 days after 1-100 D-D pulses is 64 Cu (half life 12.7
hr), and later on - 60 Co (half life 5.27 yr) produced as a result of the 63 Cu (n, α) 60 Co - reaction
caused mainly by the D-T neutrons. An impact of the D-D neutrons on the total dose rate is
negligible.
In case of a longer cooling after 1000 D-D pulses, the radionuclides in irradiated steel
dominate. 56 Mn (2.58 hr), 57 Ni (36.1 hr), and 58 Co (70.8 days) are the main contributors to
the dose rate 1-30 days after shut down, and 54 Mn (312.5 days) and 60 Co (5.27 yr) - after
cooling one year. An impact of the D-D source neutrons on the total dose rate for steel is only
about ~20% of the total value in the periods one day to one year after reactor shutdown.
The results show that under the D-D operation the plasma chamber activation is three orders
of magnitude lower than expected after the D-T phase. Nevertheless, a dose rate ~1000
µSv/hr exceeding the hand-on maintenance limit ~100 µSv/hr, will be maintained for several
1 R. A. Forrest and J-Ch. Sublet, FISPACT 97: User Manual, The European Activation System (EASY-97),
EASY Documentation Series, UKAEA FUS 358, UKAEA Fusion, Culham, Abingdon, May 1997.
2 R. T. Santoro, V. Khripunov, H. Iida et al., ITER Nuclear Analysis Report. G 73 DDD 1 98-06-17 W0.2,
NAG-101-98-06-17-CDR, Garching, June 1998.

ITER G 73 DDD 2 01-06-06 W0.1
months after a single 300-s DD-pulse. Personnel access to the plasma facing structures, such
as the first wall and the divertor targets, will be restricted already after only a few tens of
seconds of “full scale” DD-operations.
At the same time the radiation conditions outside the machine behind the bulk radiation
shield, where a dose rate ~1 µSv/h is expected, will allow for hands-on-maintenance almost
immediately after reactor shutdown.
Residual γ-energy release in the first wall and blanket materials is a function of the total
neutron flux during a D-D pulse and the total neutron fluence. Specific decay γ-heating in the
Cu- and SS- layers of first wall does not exceed ~1.3 - 3 10 –4 W/cc. It is about two orders of
magnitude lower than during the D-D-plasma burn (See Table 8-4). A total afterheat ~6 - 9
kW is expected immediately after reactor shutdown at the end of the D-D phase.
8.6 Concluding remark
The analysis showed that many systems initially intended for the nominal D-T operations
(such as tritium production and removal, active cooling and heat rejection during -plasma
burn and after shut down, as well as remote handling of the in-vessel components, neutron
diagnostics, and many ancillary systems) will need to be available from the very beginning to
realise the advanced D-D operations. For this reason the foreseen D-D operations can be
treated as an initial nuclear phase. So the presented nuclear performance estimates seem to be
important for further considerations of radiation safety, maintenance and staged ITER
construction.
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ITER G 73 DDD 2 01-06-06 W0.1
9 Uncertainty Estimate
The estimations described in the above chapters are the ‘best estimates’ and do not include
any safety margins. It is necessary to have some idea about the uncertainties those estimates
may have. Especially, we should have such information on the subjects, which are critical for
the machine design, such as nuclear heating in the magnets, helium production rate in the insitu
welding position and dose rates around the machine after shutdown.
The uncertainties may consist of the following two kinds;
• uncertainty coming from the reliability of tools (calculation codes and nuclear data);
• uncertainty coming from deference between real design and calculation model geometry,
materials and operation condition of the machine.
The first kind uncertainty can be investigated quantitatively by conducting benchmark
experiments. The second one is more difficult to assess, since it depends on how much efforts
can be put in the actual design analyses with available man power and computer
performance.
In the following sections, the results of benchmark experiments are reported. Section 9.1
briefly summarises the results of experiments for determining uncertainties of nuclear
responses in operating condition, which has been already reported in Nuclear Analysis Report
for ’98 FDR-ITER 1 . The sections 9.2 and 9.3 describe the results of new experiments for
investigating accuracy in shutdown dose calculation. In the concluding section 9.4, overall
uncertainties are discussed with some empirical assessment for the second kind uncertainties.
9.1 Summary of T16, T218, T362 Experiments
In order to verify analytic tools and nuclear data, a series of benchmark experiments
(T16,T218 and T362) were conducted in the framework of neutronics R&D tasks. The
experiments employed mock-ups of stainless steel and water mixture which is the major
material of radiation shielding in ITER machine. T16 is the basic and bulk shielding
experiment in which shield thickness of ~1 m was selected fully covering ITER inboard
shield thickness. Consequent experiments were conducted changing geometry from simple to
more complex one with radiation streaming through penetrations.
For the analytic tools, our focus was placed on the MCNP for radiation transport code and
FENDL1 and 2 for nuclear data, since they are main tools employed in our actual nuclear
analysis.
Summary of results are shown in Table 9.1-1. Nuclear heating and fast neutron flux are
predicted with the accuracy of ± 30% by using above mentioned tools (this does not
necessary mean that accuracy become worse with other tools). As far as actual geometry can
1 R. T. Santoro, V. Khripunov, H. Iida G 73 DDD 1 98-06-17 W0.2, NAG-101-98-06-17-CDR, “ITER
NUCLEAR ANALYSIS REPORT”
Nuclear Analysis Report Page 178

ITER G 73 DDD 2 01-06-06 W0.1
be exactly simulated by using 3-D Monte Carlo code MCNP, complexity of geometry does
not reduce accuracy of calculation results. Generally calculated nuclear heating and threshold
reaction rates (then also helium production rate probably) have smaller values than
experimental values.
Table 9.1-1 Results of ITER-EDA Neutronics Experiments
Issues Bulk Shielding Gap Streaming Straight Channel
Streaming
Experiments
NO.
T16, T218 T218 T362
Results of At V.V. surface: ~15 Existence of gap does Existence of straight
Comparison %
not give significant channel does not give
±(Exp.-Cal) At the first layer of increase of uncertainty significant increase of
/Exp. W.P.: ~30 % as far as calculation uncertainty as far as
uses Monte Carlo calculation uses
Codes
Monte Carlo Codes
9.2 T426 Experiment at FNG
Neutronics experiments are important to validate nuclear analyses for ITER which are based
on code and nuclear data with inherent uncertainties. An experiment was performed at the 14
MeV Frascati Neutron Generator (FNG) on a stainless steel/water assembly, in which a
neutron spectrum was generated similar to that occurring in the ITER vacuum vessel. The
mock-up was irradiated at FNG for sufficiently long time to create a level of activation which
was, after shut down, followed by dose meters for a cooling time assumed to be required for
allowing personal access. The experiment, performed in collaboration between ENEA, TUD
and FZK, was then analysed using a rigorous, two-step method, i.e. using MCNP-4-B and
FISPACT codes, and a new, one-step method with an ad hoc modified version of MCNP
used in the nuclear analysis of ITER. FENDL-2 nuclear data libraries were used in both
cases.
9.2.1 Experiment set up
The experimental assembly consisted of a block of stainless steel AISI316 and water
equivalent material (Perspex) with total thickness of 714 mm, and a lateral size of 1000 mm x
1000 mm (Figure 9.2-1). A cavity was realised within the shield assembly (126.0 mm in the
beam direction, 119.8 mm high) behind about 224.7-mm-thick shield. A void channel (27.0
mm inner diameter) was included in front of the cavity to study the effect of streaming paths
in the bulk shield. Two experimental campaigns were performed at FNG:
1. In May 8-10, 2000 the mock-up was irradiated for a total of 18 hours in three days.
1.815x10 15 neutrons (14 MeV) were produced in total. The following measurements were
carried out in the cavity by ENEA team starting from immediately after the irradiation up to
the end of July, 2000: dose rate measurements by an active dosemeter, gamma-ray dose
distribution by TLD, measurement of the Ni-58(n,p)Co-58 and Ni-58(n,2n)Ni-57 reaction
rates during irradiation by Ni activation foils.
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ITER G 73 DDD 2 01-06-06 W0.1
2. In August 29-30, 2000 the assembly was irradiated a second time, for a total of 13 hours in
two days. The total 14 MeV neutron yield was 1.95x10 15 neutrons. Measurements of the
decay gamma-ray flux spectrum and of the gamma-ray dose rate were simultaneously carried
out in the cavity by the TUD team from immediately after the irradiation up to September 21,
2000. The spectrum of the fast neutron flux in the centre of the cavity during irradiation of
the mock-up with 14 MeV neutrons was measured too.
9.2.2 Measurements
In the first experimental campaign, a continuous measurement was taken of the dose rate in
the cavity centre after shut down from half an hour to more than two months of cooling time,
using a Geiger-Muller detector (G-M, Mod. 7312 - Vacutec) with a Multi-Channel Scaler
with variable dwell time (EG&G Ortec). The detector (12 mm in diameter, 80 mm in length)
was located in the cavity centre in front of the open channel (Figure 9.2-1). The total
experimental uncertainty was ± 10% for G-M detector.
High sensitivity thermoluminescent detectors of the type TLD-300 (CaF2:Tm) GR-200A
(LiF:Mn, Cu, P) were also used to measure independently the dose rate in the cavity centre
(close to G-M) and around the cavity walls (Figure 9.2-1) at four decay times (8.2, 12.4, 19.2
and 33.2 days, for time intervals ranging from 18 to 22.5 hours). The total error associated
with the measurements was ±17%. The dose rates measured with TLD in the cavity centre
was in agreement within 12% with values obtained with the Geiger-Muller detector, within
the combined experimental uncertainties.
Figure 9.2-1 Schematic view of experimental set-up (vertical cut) with the
positions of the various detectors during the two experimental campaigns
Activation measurement were carried out using Ni foils located on the cavity walls (Figure
9.2-1). The goal was to measure the reaction rate of Ni-58(n,p) producing the Co-58
(responsible of most of the dose rate in the relevant decay time), and the reaction rate of Ni-
58(n,2n) which produces the Ni-57 (the second most important contributor to total dose rate
in the first week after shutdown, after Mn-56 is decayed) (Figure 9.2-2). The total
experimental error was ±5%.
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ITER G 73 DDD 2 01-06-06 W0.1
The dose rate in the cavity was measured in the second campaign using a tissue-equivalent
scintillator (46 mm in both, diameter and height; mab 500 modified dose meter) with high
sensitivity (range of measurement from 0.05 μSv/h to 10 mSv/h). It was calibrated at PTB
Braunschweig in standard photon fields with energies ranging from E γ =19.9 keV up to E γ =
6.7 MeV. The uncertainty is 1.7%. (Figure 9.2-1)
The neutron flux spectrum was measured with a cylindrical NE213 scintillator (diameter and
length: 3.8 cm) used already for the ITER tasks T-218 and T-326. Response matrix and data
evaluation procedure are described in the corresponding reports.
Photon flux spectra were measured with the same NE213 spectrometer (Figure 9.2-1) at t =
2.08 h, 15.9 h, 25.2 h, 4.0 d, 8.2 d, 12.2 d and 19.3 d after the end of the irradiation.
The background dose rate level was accurately measured inside the block cavity: a very
steady value of 0.33 μSv/h was found before the first campaign in May, and 0.5178 μSv/h
before the second campaign in August.
9.2.3 Experiment analysis and results
The experiment analysis was performed using two different approaches:
a rigorous two-step method (R2S) employing the MCNP-4C code with FENDL/MC-2.0 cross
sections for calculating neutron and decay gamma transport in sequential order and the
FISPACT inventory code with FENDL/A-2.0 activation cross sections for calculating the
decay gamma source distribution as a function of irradiation history and cooling time;
a direct one-step (D1S) method with a modified version of MCNP code using ad hoc libraries
in which the activation cross sections are taken from FENDL-2.0; both FENDL/MC-2.0 and
FENDL/A-2.0 were used with this method. Decay gamma energies and yields were
taken from EAF-99. The activation of the following, most relevant radioisotopes was
considered: Mn-56, Co-58, Ni-57, Mo-99, Mn-54, Cr-51, Fe-59 and Co-60.
Fraction of total dose rtae (%)
100
90
80
70
60
50
40
30
20
10
0
1.0E-03 1.0E-02 1.0E-01 1.0E+00
Cooling time (years)
Figure 9.2-2 Contributions to the total dose rate of radioisotopes considered in the D1S
method, in a cell at the cavity front wall, versus cooling time. The sum is given by the black
line.
Dose rate in the cavity centre by G-M dosemeter
Mn-56
Ni-57
Co-58
Mo-99
Cr-51
Mn-54
Fe-59
Co-60
SUM
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ITER G 73 DDD 2 01-06-06 W0.1
The measured values are compared with the results of R2S calculations in Figure 9.2-3 (Δ)
for cooling times from 1 day to 2 months. The statistical uncertainty is ≤ 2% in the
calculation of both neutron fluxes and gamma dose rates. The analysis was carried out also
with the D1S method for the same cooling times; the production rate of radioactive nuclides
was calculated using FENDL/A-2.0 (O in Figure 9.2-3) and FENDL/MC-2.0 ( in Figure
9.2-3). The statistical uncertainty ranges between 5 and 10%. All results are given in Tables
9.2-1a and 1b together with C/E ratios. The total uncertainty on the comparison is obtained
summing by quadratic law the uncertainty on E (±10%), on C (given in the tables), and that
on FNG source calibration (±3%).
Table 9.2-1a Comparison between measured and calculated dose rates in the
cavity centre.
Decay time E
(Sv/h)
R2S (FENDL/A-2)
days years C (Sv/h) C/E
1 2.74E-03 2.46E-06 ± 10% 2.60E-06 ± 3% 1.06 ± 0.12
7 1.92E-02 6.99E-07 ± 10% 7.20E-07 ± 3% 1.03 ± 0.11
15 4.11E-02 4.95E-07 ± 10% 5.37E-07 ± 3% 1.08 ± 0.12
30 8.22E-02 4.16E-07 ± 10% 4.61E-07 ± 3% 1.11 ± 0.12
60 1.64E-01 3.16E-07 ± 10% 3.51E-07 ± 3% 1.11 ± 0.12
Table 9.2- 1b Comparison between measured and calculated dose rates in the
cavity centre.
Decay time D1S (FENDL/A-2) D1S (FENDL/MC-2)
days years C (Sv/h) C/E C (Sv/h) C/E
1 2.74E-03 1.62E-06 ± 9% 0.66 ± 0.09 2.13E-06 ± 10% 0.87 ±0.13
7 1.92E-02 5.77E-07± 5% 0.83 ± 0.10 6.56E-07 ± 6% 0.94 ±0.11
15 4.11E-02 4.32E-07 ± 5% 0.87 ± 0.10 4.50E-07 ± 6% 0.91 ±0.11
30 8.22E-02 4.12E-07 ± 5% 0.99 ± 0.12 4.23E-07 ± 6% 1.02 ±0.12
60 1.64E-01 3.18E-07 ± 5% 1.00 ± 0.12 3.27E-07 ± 6% 1.03 ±0.12
In the R2S case the calculated values of dose rate are in agreement with the measurements
within the total uncertainty on the comparison. All C/E values between 1 day and 2 months of
decay time are close to unity within 11%. The D1S method gives values systematically lower
than the R2S method, especially for decay times ≤ 15 days. In part (≈10 %) this may be due
to the fact that minor radioisotopes, each contributing to the total dose rate at the percent
level, are not considered in the D1S method (see Figure 9.2-2).
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ITER G 73 DDD 2 01-06-06 W0.1
1.E-05
1.E-06
1 day
7 days
15 days
Measured Doserate (G-M)
Measured Doserate (TLD)
R2S (FENDL-2/A)
D1S (FENDL-2/MC)
D1S (FENDL-2/A)
1 month
2 months
1.E-07
1.E-03 1.E-02 1.E-01 1.E+00
Time after irradiation (y)
Figure 9.2-3 Comparison of measured and calculated dose rate in the cavity centre
Dose rate in the cavity by dose meter with tissue-equivalent scintillator
The dose rate measured with the tissue-equivalent scintillator is compared in Figure 9.2-4a
with results of R2S and D1S calculations using FENDL-2/A cross-section data. As a general
trend, there is observed an overall satisfactory agreement over the considered range of
cooling times. A more detailed C/E (calculation/ experiment) comparison reveals an
underestimation of the measured dose rate at short cooling times (less than ≅ 2 days) by up to
15 % in case of the R2S calculation (Figure 9.2-4b). The D1S-calculation underestimates the
measured dose rate by 20 to 25 %. The dominating dose rate contributions are due to 56 Mn
(few hours cooling time) and 57 Ni (1 – 5 days cooling time). At larger cooling times, the 58 Co
contribution gives rise to an overestimation of up to 10%.
Neutron flux spectrum during irradiation
The measured flux spectrum is compared in Figure 9.2-5 with the result of a MCNP-4B/
FENDL-2 calculation. The dominant part at the detector position on the streaming channel
axis is the 14-MeV neutron peak. It contains 71% of the neutron flux with E > 1 MeV, and
with increasing E more and more reaction channels are open and produce radioactivity. The
measured fluence of neutrons with E > 13.7 MeV, normalised to one source neutron amounts
to (5.91 ± 0.35)·10 -5 . The corresponding calculated value is (5.96 ± 0.32)·10 -5 , resulting in a
ratio of calculated-to-experimental fluence (C/E) of 1.01 ± 0.07. (Total uncertainty for E and
statistical one for C)
Nuclear Analysis Report Page 183

ITER G 73 DDD 2 01-06-06 W0.1
Dose Rate (mSv/h)
100
10
1
10 4
Nuclear Analysis Report Page 184
10 5
Cooling time [s]
Measurement (tissue-equ. scintillator)
R2S calculation (FENDL-2/A+MC)
D1S calculation (FENDL-2/A)
Figure 9.2-4a Comparison of calculated and measured (tissue-equivalent
scintillator) dose rates.
10 6

ITER G 73 DDD 2 01-06-06 W0.1
Calculation / Experiment
1,20
1,15
1,10
1,05
1,00
0,95
0,90
0,85
0,80
0,75
0,70
0,65
0,60
10 4
R2S calculation (FENDL-2/A + MC)
D1S calculation (FENDL-2A)
Experimental uncertainty
Nuclear Analysis Report Page 185
10 5
Cooling time [s]
Figure 9.2-4b C/E (calculation/experiment) comparison for the dose rate measured
with the tissue-equivalent scintillator.
-1 -
cm
2 )
( MeV
Neutr on fl uen ce
1.0E-3
1.0E-4
1.0E-5
1.0E-6
1.0E-7
1.0E-8
____ Measurement
Calculation
0.0 4.0 8.0 12.0 16.0
Neutron energy (MeV)
Figure 9.2-5 Comparison of measured and calculated fast neutron fluence,
normalised to one source neutron
10 6

ITER G 73 DDD 2 01-06-06 W0.1
Decay gamma ray spectra in the cavity
)
-1 -2 -1
Decay gamma flux density (MeV
*cm *s
1,8x10 3
1,6x10 3
1,4x10 3
1,2x10 3
1,0x10 3
8,0x10 2
6,0x10 2
4,0x10 2
2,0x10 2
0,0
t=15.9h
Experiment (shaded)
R2S calculation (FENDL-2/A + MC)
D1S calculation (FENDL-2/A)
1 2 3
Gamma energy (MeV)
Figure 9.2-6a Comparison of measured and calculated decay
gamma ray spectra at 15.9h after irradiation.
-2 *s -1
10000
)
Decay gamma flux (cm
1000
100
10
10 4
Nuclear Analysis Report Page 186
10 5
Cooling time [s]
R2S calculation total
D2S calculation total
R1S calculation E > 0.4 MeV
D1S calculation E > 0.4 MeV
TUD measurement E > 0.4 MeV
Figure 9.2-6b Comparison of calculated and measured decay
gamma fluxes
10 6

ITER G 73 DDD 2 01-06-06 W0.1
Calculation / Experiment
1,3
1,2
1,1
1,0
0,9
0,8
0,7
0,6
0,5
10 4
R2S calculation (FENDL-2/A+MC)
D1S calculation (FENDL-2/A)
Experimental uncertainty
Nuclear Analysis Report Page 187
10 5
Cooling time [s]
Figure 9.2-6c C/E (calculation/experiment) comparison for the measured decay gamma
fluxes.
The measured decay gamma ray spectra can be well reproduced by the calculation, see e. g.
Figure 9.2-6a for the spectra at 15.9 h after irradiation. The D1S calculation shows a rather
fine resolution of the gamma peaks because discrete gamma lines are included in the decay
gamma spectra of the ad-hoc prepared data library. The R2S-calculation handles the decay
gamma spectra in a 24 group structure as provided by the FISPACT code.
For gamma energies above 0.4 MeV, measured and calculated decay gamma fluxes are
compared in Figure 9.2-6b as a function of cooling time. A corresponding C/E comparison is
displayed in Figure 9.2-6c. There is the same trend as for the dose rate, except for the
shortest cooling time. The C/E >≈ 1 for t = 2.08 h comes from overestimation in the lowenergy
part of the spectrum, whereas the high-energy part is slightly underestimated. But the
dose rate is determined by the flux spectrum increasingly with energy.
Dose rate distribution along the cavity walls
The dose rate distribution was measured by TLD detectors on the cavity walls. The results
were analysed by the R2S method (statistical uncertainty ≤3%) and with the D1S method
(statistical uncertainty in the range 3% - 8%). The comparison of R2S and of D1S methods
show a general agreement within ±20%, with the exception of the case at 1 day of cooling
time, where the R2S result is about 25% higher than the D1S + FENDL/A-2.0 one,
consistently with the results obtained in the calculation of the dose rate in the cavity centre.
The comparison with the experimental data shows that the global features of the dose rate
local distribution inside the cavity are satisfactorily predicted by calculation.
10 6

ITER G 73 DDD 2 01-06-06 W0.1
Co-58 and Ni-57 production rate by activation measurements
The neutron flux at the Nickel foils during irradiation was calculated using MCNP; the Ni-
58(n,p)Co-58 or Ni-58(n,2n)Ni-57 reaction rates were calculated in two ways:
1. using a procedure similar to R2S method, i.e. using FISPACT with Ni-58(n,p) and (n,2n)
cross sections from FENDL/A-2. Statistical errors on MCNP flux calculations are
≤±2.5%.
2. using a procedure similar to D1S method in which the reaction rate is directly calculated
in the MCNP run taking the Ni-58(n,p) and (n,2n) cross sections from the dosimetry file
IRDF-90.2 and from FENDL/MC-2. Statistical errors on reaction rate calculations are ≤
±2.5%.
All results for Ni-58(n,p) reaction are given in Table 9.2-2. The total errors on C/E are ± 0.05
and include the statistical error in the MCNP calculations and the experimental error on
measurement (±4.5%). Using FISPACT with FENDL/A–2 (R2S), C/E values are in general
slightly higher than unity, while the C/E values obtained with direct MCNP calculation (D1S)
using the dosimetry file IRDF-90 and FENDL/MC–2 are generally lower, and in better
agreement with the measurements. These results are coherent with those found in the analysis
of the dose rate by R2S and D1S methods.
As far as the Ni-58 (n,2n) reaction is concerned, all results are given in Table 9.2-3. In the
D1S calculation with the dosimetry file IRDF-90 and FENDL/MC–2, C/E values in average
are slightly lower than unity, but still within the total uncertainty (±5%). Using FISPACT
with FENDL/A–2 (R2S), C/E are lower than in the previous case by about 1-3%.
Table 9.2-2 Measured and calculated Ni-58(n,p)Co-58 reaction rates (x10 -24 /source
neutron).
Foil Meas. MCNP +FISPACT MCNP Calculation
number E C (A-2) C/E C (IRDF-90.2) C (MC-2) C/E
1 2.15E-5 2.196E-5 1.02 2.099E-5 2.099E-5 0.98
2 5.19E-6 5.511E-6 1.06 5.447E-6 5.446E-6 1.05
3 4.13E-6 4.558E-6 1.10 4.022E-6 4.021E-6 0.97
4 8.48E-6 9.299E-6 1.10 8.637E-6 8.637E-6 1.02
5 7.86E-6 8.311E-6 1.06 7.697E-6 7.695E-6 0.98
6 5.15E-6 5.345E-6 1.04 4.990E-6 4.899E-6 0.97
Table 9.2-3 Measured and calculated Ni-58(n,2n)Ni-57 reaction rates (x10 -24 /source
neutron).
Foil
number
Meas.
E
MCNP+FISPACT
C (/A-2) C/E C
MCNP Calculation
C
C/E
(IRDF-90.2) (MC-2) 90.2 / MC2
1 2.84E-6 2.597E-6 0.91 2.618E-6 2.653E-6 0.92 / 0.93
2 3.94E-7 3.947E-7 1.00 4.029E-7 4.069E-7 1.02 / 1.03
3 2.07E-7 1.912E-7 0.92 1.943E-7 1.965E-7 0.94 / 0.95
4 4.92E-7 4.991E-7 1.01 5.105E-7 5.158E-7 1.04 / 1.05
5 4.71E-7 4.282E-7 0.91 4.363E-7 4.408E-7 0.93 / 0.94
6 3.64E-7 3.273E-7 0.90 3.339E-7 3.374E-7 0.92 / 0.93
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ITER G 73 DDD 2 01-06-06 W0.1
9.2.4 Conclusions
• A mock-up was irradiated at FNG with 14 MeV neutrons for sufficiently long time and a
high level of activation was created, which allowed to measure the dose rate for more
than two months of cooling time after shut down, using two independent experimental
techniques. Other measurements useful for the analysis were performed, such as the nspectrum,
the decay γ-ray spectrum, the dose rate distribution and some relevant
activation reaction rates.
• The experiment, was analysed using a rigorous, two-step method (R2S), i.e. using
MCNP-4-B and FISPACT codes, and a direct, one-step method (D1S) with an ad hoc
modified version of MCNP used in the nuclear analysis of ITER. FENDL-2 data libraries
were used.
• The dose rate measurement with Geiger-Muller detector is well predicted by R2S method
within the total uncertainty on the comparison (±11%): all C/E values between 1 day and
2 months of decay time are close to unity within 11%. The D1S method, when using the
same cross section file (i.e. FENDL/A-2.0), is also in good agreement with measurements
and gives values slightly but systematically lower than the R2S method.
• The dose rate measured with the tissue-equivalent scintillator agree within ±15% with
R2S calculations using FENDL-2/A cross-sections. There is a tendency for
underestimating the dose rate at short decay times (≤ ≅ 2 days) which can be addressed to
Mn-56 and Ni-57. This tendency is enhanced for the D1S method in which minor
nuclides, contributing to the total dose rate at the percent level, are not considered. At
larger cooling times, the dose rate is slightly overestimated. This is likely due to the Co-
58 production cross-section in FENDL-2/A.
• For the Ni-58(n,p)Co-58 activation measurements the R2S method gives C/E values
slightly higher than unity, while the C/E values obtained with the D1S method using
IRDF-90 and FENDL/MC–2 are in better agreement with the measurements. For the Ni-
58(n,2n)Ni-57 reaction, both methods give C/E values slightly lower than unity, the
underestimation being more pronounced by D1S. These results are coherent with those
found in the analysis of the dose rate by the R2S and D1S methods.
• The measured decay γ−ray spectra can be reproduced rather well by the calculations. The
resolution of the calculated decay gamma peaks, however, depends on the way the
underlying decay gamma emission spectra are represented in the calculation. The D1S
calculation makes use of discrete decay gamma lines while the R2S method employs the
24 group structure as provided by FISPACT. Discrete decay gamma emission spectra,
however, may be applied in the R2S calculation as well.
• The decay gamma flux, measured above 0.4 MeV photon energy in the experiment, can
be calculated with an uncertainty of ±15% when using FENDL-2/A cross-sections. As
with the dose rate, there is a tendency for underestimating the measurements, in particular
in the time range 1 to 5 days after irradiation. Again this underestimation is larger for the
D1S method than it is for the R2S method.
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ITER G 73 DDD 2 01-06-06 W0.1
9.3 T426 Experiment at FNS/JAERI
In the shielding design of the ITER machine, it is important to have a reliable estimation of
the dose rate levels after reactor shutdown for realising hands-on maintenance around the
torus. In order to assure the hands-on maintenance possibility inside the cryostat without
excessive shielding, a very higher accuracy of shutdown dose estimate is required.
Credibility of nuclear design calculations requires stringent verification by comparisons with
experimental data. In this section, the results of the task T426, conducted Japanese home
team, is reported clarifying the credibility concerning the methodologies for shutdown dose
estimates used in the ITER, which is called the new ‘direct one-step method’ developed by
JCT and EU Home Team (see Appendix B).
9.3.1 Experiment set up
The experiment was performed at the Fusion Neutronics Source (FNS) facility of the Japan
Atomic Energy Research Institute (JAERI). The D-T neutrons were produced by bombarding
a water-cooled tritium-titanium target with a 350 keV deuterons beam.
The experimental assembly consisted of the source reflector of 200 mm in thickness and test
region which had a cylindrical shape of 1200 mm in diameter and 286 mm in thickness as
shown in Figure 9.3-1. The test region simulated a maintenance area and was composed of
type 316 stainless steel (SS-316). The source region and the test region were connected with
cylindrical duct of 200 mm in diameter surrounded by SS-316 and water layers, which is a
standard shielding structure of the ITER. Circles and squares show measurement positions of
shutdown dose. The neutron spectrum at the positions indicated as circles include direct
14MeV component and the ones at the positions indicated as squares were shielded from D-T
target by SUS316/water shield.
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ITER G 73 DDD 2 01-06-06 W0.1
Ref lect or
(SS316)
800
300
200
Shield
(SS316/ H20)
#a #b #c
#A#B #C
356
Measurement
posit ion
1200
D-T source Test region Ref lect or (SS316)
Figure 9.3-1 Schematic view of experimental set-up with the positions of the
measurement
The irradiation was conducted through 6 days. Total and average neutron intensity were 2.1
x10 16 neutrons and 9.6x10 10 neutrons per second respectively. The irradiation schedule was
determined by pre-analysis so that the shutdown dose rate could be measured at 10 6 seconds
after the irradiation.
9.3.2 Measurement
Various experimental data were measured mainly in the test region. The measurement items
and methods are as follows; i) shutdown dose rate from 2.2x10 5 s (2.5 day) to 1.44 x10 6 s
(16.6 day) after irradiation with a tissue equivalent dose meter, ii) decay gamma spectrum
with a scintillation spectrometer, iii) neutron spectra above 2 MeV with a scintillation
spectrometer, iv) dosimetry reaction rates of Nb, In and Au with the foil activation method,
and v) fission rates of 235 U and 238 U with micro fission chambers.
9.3.3 Experiment analysis and results
In the shielding development of the ’98 FDR-ITER design, the conversion factor method was
used to evaluate shutdown dose and the conversion factor (from fast neutron flux to
shutdown dose rate) was estimated by the full step ( neutron transport – activation calculation
– gamma transport) calculation in a simple geometry. Though the conversion factor method
is convenient for rough and quick estimation of shutdown dose since only fast neutron flux is
required to be calculated, the design margin should be large because the uncertainty
accompanied with results may not be small.
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ITER G 73 DDD 2 01-06-06 W0.1
On the other hand, the new direct one step method employs 3-dimensional Monte Carlo
transport code MCNP for calculating decay gamma-ray transport and provides shutdown
dose rate directly without using conversion factor. Therefore this method can eliminate
uncertainty accompanied with conversion factors. Some details of this method are given in
Appendix B. The nuclear data libraries used for neutron transport are FENDL/1 MC, and
FENDL/2 MC. The activation cross sections, which produce main contributing radioisotopes,
are taken from those included in FENDL/1 MC and /2 MC and from FENDL/2A.
Calculation results of the radio-activities of isotopes with the actual irradiation history in the
experiment are shown in Figure 9.3-2. The main isotopes contributing to the shutdown dose
rate at around 10 6 sec after the irradiation are 58 Co, 51 Cr, 54 Mn, 59 Fe and 99 Mo. Some of these
isotopes have significant contributions to the shutdown dose of the ITER and decay gamma
rays of these are considered in the shielding design. The measured decay gamma ray
spectrum at the position along the axis and the peripheral at 8.5x10 5 (sec) are shown in Figure
9.3-3, and the decay gamma rays of these nuclides were observed.
Comparison of the measured and calculated tissue equivalent dose rate at around 10 6 sec after
the irradiation are shown in Figure 9.3-3. The C/E values of the shutdown dose rates are
summarised in Table 9.3-1and Figure 9.3-4. The statistical uncertainty is less than 4% in the
calculation of both neutron flux and gamma dose rate. The production rates of radioactive
nuclides were calculated using FENDL/1, FENDL/2 and FENDL/2A. The total uncertainty
of the comparison is obtained summing the uncertainty on E, on C, and that on source
calibration (~3%). The results with FENDL/2 and FENDL/1 are almost the same, because
nuclear data for main isotopes contributing to the shutdown dose are the same for these
libraries. On the other hand, shutdown dose rates evaluated with FNEDL/2A were slightly
larger than those with FENDL/1 and FENDL/2. Contributions to the total dose rate of
radioisotopes at 1e+6 sec after the irradiation are shown in Figure 9.3-5. As shown in this
figure, the difference between the results of FENDL/2 and FENDL/2A, because the evaluated
cross section of 58Ni(n,p) reaction in FENDL/2 is larger than the others.
1.0E+12
1.0E+11
1.0E+10
58m C
57 Ni
1.0E+09
2.0E+05 4.0E+05 6.0E+05 8.0E+05 1.0E+06 1.2E+06 1.4E+06 1.6E+06
Cooling t ime
Tot al
Nuclear Analysis Report Page 192
51 Cr
58 Co
99m Tc
99 Mo 54 Mn
59 Fe
Figure 9.3-2 Calculation results of radioactivity according to the irradiation profile

ITER G 73 DDD 2 01-06-06 W0.1
1E+4
1.0E+04
1E+3
1.0E+03
1E+2
1.0E+02
1E+1
1.0E+01
1E+0
1.0E+00
1E-1
1.0E-01
51C r
0.511
58C o
54 Mn
60C o
59 Fe
60C o
59 Fe
58C o
57 Ni
0.0 0.5 1.0 1.5 2.0
0.0 0.5 1.0 1.5 2.0
Gamma ray energy ( MeV)
Figure 9.3-3 Measured decay gamma ray spectrum after 8.5x10 5 sec after the
irradiation
Table 9.3-1 (a) C/E value of shutdown dose rate evaluated by
FENDL/2
Coolin time
(sec)
#A #a #B #b #C #c
2.20E+05 0.96Å}0.10 0.90Å}0.09 0.94Å}0.10 0.93Å}0.09 0.83Å}0.09 0.91Å}0.09
5.08E+05 0.97Å}0.11 1.01Å}0.10 0.99Å}0.12 1.03Å}0.10 0.91Å}0.11 1.01Å}0.10
8.30E+05 0.96Å}0.11 1.05Å}0.10 1.03Å}0.12 1.06Å}0.10 0.92Å}0.12 1.03Å}0.10
1.44E+06 1.00Å}0.10 1.03Å}0.10 1.04Å}0.13 1.09Å}0.11 0.96Å}0.12 1.07Å}0.11
Table 9.3-1 (b) C/E value of shutdown dose rate evaluated by
FENDL/2A
Coolin time
(sec)
#A #a #B #b #C #c
8.30E+05 1.04Å}0.10 1.15Å}0.10 1.06Å}0.11 1.18Å}0.10 1.03Å}0.11 1.14Å}0.10
1.44E+06 1.07Å}0.11 1.12Å}0.11 1.07Å}0.11 1.21Å}0.11 1.07Å}0.11 1.18Å}0.12
Nuclear Analysis Report Page 193

ITER G 73 DDD 2 01-06-06 W0.1
9.0
8.0
7.0
6.0
5.0
4.0
Dose rate
(μSv/hr)
3.0
2.0
1.0
0.0
0.5
1.0E+05 3.0E+05 5.0E+05 7.0E+05 9.0E+05 1.1E+06 1.3E+06 1.5E+06
Figure 9.3-4 Time dependence of measured and calculated tissue
equivalent dose rate
Shutdown dose_μSv/hr_
5.0
_
4.0
_
3.0
_
2.0
_
1.0
_
0.0
5.0
4.5
4.0
3.5
3.0
2.5
2.0
1.5
1.0
0.5
0.0
Position #A
FENDL/2
FENDL/2A
Cooling time
FEND
FENDL/
Experime
Position #a
Co58
58Co
Co60 60Co Fe59 59Fe Mn54 54Mn Cr51 51Cr Mn56 56Mn Mo99 TOTAL
99Mo TOTAL
Isotope
Figure 9.3-5 Contributions to the total dose rate of radioisotopes at 10 6 sec after the
irradiation
Nuclear Analysis Report Page 194
3.5
3.0
2.5
2.0
1.5
1.0
Dose rate (μSv/hr)

ITER G 73 DDD 2 01-06-06 W0.1
9.3.4 Conclusions
Through these analyses, the following conclusions can be derived.
• Shutdown dose rates evaluated by the new direct one step method with FNEDL/2 library
agreed with the experimental results with the tissue equivalent dose meter within the
experimental error (~10%). The accuracy of the shutdown dose evaluation by the new
direct one step method with FENDL/2 is small enough, comparing with the target
accuracy of the factor two.
• Main isotopes which contributes the shutdown dose rate at 10 6 sec after irradiation were
similar as those considered in the current ITER design methodologies.
• Shutdown dose rates evaluated with FNEDL/1 were almost the same as the results with
FENDL/2 library, because the basic nuclear data for the main isotopes are the same for
both libraries.
• Shutdown dose rates evaluated with FNEDL/2A were slightly larger than those with
FENDL/1 and FENDL/2 library, because the evaluated cross sections of 58Ni(n,p)
reaction in FENDL/2 are larger than the others.
9.3.5 Summary
In order to clarify the required safety factors concerning the methodology for shutdown dose
estimates (one step Monte Carlo method) developed by JCT and EU Home Team, the
benchmark experiment for the shutdown measurement were conducted at FNS/JAERI. The
results showed that the accuracy of the one step Monte Carlo method with FENDL/2 library
is sufficient for the shielding design of the ITER.
9.4 Discussion
As previously stated, the various estimates of nuclear responses given in this report are ‘the
best estimates’ and do not include any safety margin. Naturally they should have some
uncertainties. The uncertainties originate, on the one hand, from reliability of the analytic
tools and, on the other hand, come from the fact that the actual machine can not been
modelled exactly in the actual design calculation. The former can be quantified by
conducting benchmark experiments. The latter mainly comes from limitations of man power,
capacity of computer available and information on the details of machine design and is not
easy to quantify.
In this section, some discussion is developed focusing on critical nuclear responses for
determining machine shielding size and configuration. They are nuclear heating in magnets
and helium production rate at the vacuum vessel surface during operation and dose rates after
shutdown.
9.4.1 Uncertainty in nuclear heating and helium production estimate
As reported in section 9.1, it is possible to predict nuclear responses with accuracy of 15% at
the surface of vacuum vessel and of 30% at the inboard TF coil, as far as MCNP and
Nuclear Analysis Report Page 195

ITER G 73 DDD 2 01-06-06 W0.1
FENDL/1 or /2 nuclear data are used with fully detailed modelling which is possible for
benchmark experiment but generally not for the actual machine.
In addition to these intrinsic uncertainties of analytic tools, the possible elements we should
think about are listed as follows with empirical estimation of uncertainty amounts.
a) Fine structure of the blanket modules
The blanket modules are treated as a few layer of homogenised mixture in the 3-D MCNP
model. Special concerns are simple smearing of the “grooved area” at the back side of the
module and homogenisation of the radial water channels in the module. Depending on the
fine structure, integral nuclear heating in the magnet (and vacuum vessel also) will be
affected by less than 20 %. When definite design of blanket module is established, this effect
can be more definitely quantified in future.
b) Fine structure of the vacuum vessel
Heterogeneity effect of the vacuum vessel on the TF coil inboard leg was assessed and
reported in Chapter 4 (see 4.2.1 and 4.3.4). These effects (flexible joint void: 1.2 and
heterogeneity: 1.2) are already incorporated in the inboard leg heating estimate. For the
outboard nuclear heating (and/or flux), the main neutron leakage should be from the ports,
instead of passing through the blanket and bulk part of the vacuum vessel. Thinking that the
resulting flux attenuation is larger by 2 order of magnitude, heterogeneity effect of 40 % is
assumed for the outboard heating (and/or flux).
c) Maximum value of gap width among blanket modules
The nominal gap width is 2 cm, and the maximum value can be larger (up to 2.5 cm). Gap of
2.5 cm results in a peaking factor 20 % higher than that for 2 cm as shown below (Figure
9.4-1).
Peaking Factor
6
5
4
3
2
1
0
0 0.5 1 1.5 2 2.5 3
Gap Width (cm)
Figure 9.4-1 Peaking factor of fast neutron flux on the surface of the vacuum vessel
(by the Handy Method)
d) Others
Other factors, which are not easy to quantify, are
Access holes for remote handling
Diagnostics on the inboard
Nuclear Analysis Report Page 196

ITER G 73 DDD 2 01-06-06 W0.1
We take the factor 1.1 for both of local and integrated values.
Table 9.4-1 Summary of uncertainty analysis
Location Exp.-
Modelling Total
Calc. Fine Structure Max. Others
Blnkt V.V. Gap
Helium-pro. at VV
(local value) surface
Inboard 15% - - 20% 10% 1.5
Nuclear
heating
at TF coil Inboard 30% 20% - - 10% 1.7
(Integral
value)
Note:
Outboard 30% 40% - - 1.8
Fractional standard deviation in Monte Carlo analysis is usually small enough (< 5 %)
The above discussion assumes that the FENDL/1 or /2 nuclear data and MCNP are used.
9.4.2 Uncertainty in shutdown dose rate estimate
Shutdown dose rate estimate can not be done better than the above estimate which is for
during reactor operation. Shutdown dose rate requires additional steps of calculation with
using neutron fluxes obtained for operation condition. They are activation calculation and
decay gamma-ray transport calculation. Good calculation tools for those steps already exist
and an issue is how to combine them.
Combination method is straightforward, when discrete ordinate transport codes, which define
fine meshes in calculation geometry, are used for the cases of simple geometry. Simple
combination of the three steps; a neutron transport – an activation calculation in each mesh –
a decay gamma transport, works well without any problem.
However, when Monte Carlo method, which does not have fine meshes but has limited
number of cells instead, is applied, approximation of flat distribution of decay gamma source
in cells can cause significant error in dose estimation in actual design calculation.
The one step Monte Carlo method was proposed and developed by JCT and EU HT 1 23 to
solve this problem. Some details of the method are presented in Appendix B. The design
calculations have been conducted with this one step method and a two step method which
employs straightforward combination and developed EU HT 4 and RF HT 1 independently.
1 H.Iida, D. Valenza, R. Plenteda, R. T. Santoro and J. Dietz "Radiation Shielding for ITER to allow for Handson
Maintenance inside the Cryostat", J. of Nuclear Science and Technology, Supplement 1. p.235-242 (March
2000)
2 D. Valenza, H. Iida, R. Plenteda: "Proposal of Shutdown Dose Estimation Method by Monte Carlo Code" to
be published in the journal of Fusion Engineering and Design
3 L. Petrizzi, H. Iida, D. Valenza, and P. Batistoni: Improvement and benchmarking of the new shutdown dose
estimation method by Monte Carlo code. Presented at MC2000 conference Advanced Monte Carlo for
Radiation Physics Particle Transport Simulation and Applications. 23-26 Oct. 2000 Lisbon, Portugal--
4 Y. Chen and U. Fisher “ITER-FEAT Shutdown Dose Rate Analysis by Rigorous Method” EU HT report, June
2001
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ITER G 73 DDD 2 01-06-06 W0.1
Benchmark experiments were conducted with using FNG(ENEA) and FNS(JAERI) to verify
those methods and the results were described in 9.2 and 9.3. The both benchmark
experiments and their analyses provide the following results:
• At the time of 10 6 sec after shutdown and after that, the both methods agree well with the
experimental measurements within the experimental uncertainty (10%) although
calculation results tend to slightly overestimate the dose rates.
• During 2.5 days to two weeks (~10 6 sec) after shutdown, calculation results agree also
experimental results within the experimental uncertainty (10%), showing a general trend
of underestimation by the calculation when the cooling time become shorter.
• In the FNG benchmark experiment, before 2.5 days, calculation results underestimate
dose rates with larger deviation than the experimental uncertainty (up to 25 % for 1 step
method, 17% for 2 step method). The reason of this disagreement could not be identified
in this task leaving further work for future.
As a conclusion, it is enough to add further 10% of uncertainty to that for operating condition
(1.8) leaving an uncertainty factor of 2.0 for estimation of dose rate at 10 6 sec after shutdown.
This conclusion holds for the dose rate estimation during 2.5 days to 60 days (maximum of
experiments) after shutdown. A little bit larger uncertainty might be taken into account when
dose rate 2.5 days before shutdown is estimated.
The above discussion does not include statistical errors which each design calculation has.
The statistical errors of each design calculation should be added in the consideration of final
uncertainty.
In addition to the comparison using the benchmark experiments, the one step and the two step
Monte Carlo methods are compared with using an example of actual design calculation 1
(Maintenance Port Analysis). The comparison showed a factor of 2 –3 deference in their dose
rate estimates at two weeks after shutdown as expected by the JCT. The investigation and
confirmation of the reason for this difference is left for a future work.
1 Neutronic Analysis of the Vacuum Vessel/ Cryostat Environment, Report of RF HT for the 1 st quarter 2000,
CF 04-00/1,April 2000
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ITER G 73 DDD 2 01-06-06 W0.1
10 Summary of Major Nuclear Responses and
Conclusions
In the previous chapters, the results of nuclear analyses performed for ITER during the
Engineering Design Activity are given. In this section nuclear responses which have the
major effect on the machine configuration, such as nuclear heating in the superconducting
magnets and dose rate after shutdown, are collected and tabulated for deriving over all
conclusion.
10.1 Summary of Major Nuclear Responses
10.1.1 Nuclear Heating in Superconducting Magnet System
The design limit for total nuclear heating on the superconducting toroidal field coils is
specified to be 13.7 kW in the DRG1. The contributions to the integrated nuclear heating in
the toroidal field coils from radiation leaking through different reactor components are
summarised in Table 10-1. The table is followed by some clarification for each item.
The present estimate of the nuclear heating in the toroidal field coil is lower than the DRG1
recommended value but giving only small safety margin. As presented in Chapter 9 (9.4), the
value of uncertainty is estimated (Table 9.4.1). Applying the uncertainty factors of 1.7 for
inboard and 1.8 for outboard, actually possible value is calculated as follows.
10 x 1.7 + (0.84+0.45+0.285+0.38+0.3+0.1) x 1.8 =17 + 4.2
= 21.2 (kW)
The magnet system including refrigerator system should be designed so that the above
heating can be handled at the worst case by some means.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 10-1 Integrated Nuclear Heating in the Toroidal Field Coils
and Intercoil Structures
TFC Parts and Location kW Method Ref. Section
Inboard Legs ~ 10 3-D 4.5.1
Upper Part behind blanket:
(excluding “around upper ports)
~ 0.84 3-D 4.5.1
Upper Ports (18): ~0.45 3-D
(for ECH
Launcher)
5.1
Mid-Plane Ports (18): (0.285)
ICRF Ports (2) ~0.003 3-D (20° Model) 5.4
NBI Ports (3) ~0.237 3-D (60° Model) 5.2
ECH Ports (1) ~0.004 3-D (20° Model) 5.3
R. H. Ports (4) ~0.011 3-D (20° Model) 5.7
Test Blanket Ports (3) < 0.03 3-D (20° Model) 5.6
Diagnostic Ports (5) small 3-D (20° Model) 6
Divertor Ports (18): ~0.38 3-D (20° Model) 4.4.2
N16 Decay Gamma Rays: 0.3 3-D 4.5.3
around Divertor ports < 0.1
Total ~12.4
Inboard Toroidal Field Coil Legs
Nuclear heating in the inboard toroidal field coil legs was estimated based on threedimensional
analysis as reported in Section 4.5 with using ITER basic model. Some
corrections are made for detail geometry effects which can not be taken in the basic model.
They are the effect of flexible joint (1.2) and heterogeneity of the vacuum vessel (1.2) (see
sections. 4.2 and 4.3). They are included in the estimated value.
Upper Ports
Upper ports are used for plasma diagnostics or ECH for plasma control, also containing the
blanket cooling pipes along their port walls. Detailed 3-D calculations were conducted for
ECH ports for analysing nuclear responses of the launcher. The calculation model includes
shield configuration improvement at the root of the port, and gave the value of 24.5 W/port
for nuclear heating in the TF coil. It was assumed that this value can be applied for other 15
upper ports also. Improvement of shielding configuration is still under way, but the above
values should reflect this improvement fairly well.
Equatorial Ports
Detailed 3-D radiation transport analyses were conducted around the ports for plasma heating
(ICRF, ECRH, LHH and NBI) and for remote handling of blanket modules (Chapter 5). They
provide nuclear heating in the coil system around the corresponding ports.
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ITER G 73 DDD 2 01-06-06 W0.1
3-D analysis was performed for the test module ports providing the information that the
additional heating from this port will be around 10 W per port.
Equatorial port diagnostics are still being designed. 3-D analyses were conducted for some of
diagnostics system, such as, LIDAR and Polarimetry Diagnostic and Motional Stark Effect
Diagnostic Systems (Chapter 6). The analyses implies only small nuclear heating in the coil
system but absolute values were not given.
Divertor Ports
Ten of the divertor ports are for pumping out the exhaust gas and their configuration is rather
well defined. The detail of the remaining eight are still to be defined. Three dimensional
calculations were performed using the model for the pumping port providing estimate of
nuclear heating (Section 4.4). The results were applied for other 8 ports also.
16N Decay Gamma Rays
Preliminary three dimensional analyses using the 16N-gamma-source in the blanket cooling
pipe were conducted for the upper ports. Decay gamma-ray energy deposition was evaluated
in all cryogenic elements and structures in the vacuum vessel/cryostat space. (Section 4.5)
Other nuclear responses, such as specific nuclear heating on the coils, damage in copper
stabiliser and the absorbed dose in insulator are well below the DRG1 specifications.
The integral nuclear energy deposition in the poloidal field coils from neutron and secondary
photons is small, and an example of estimation (< 0.5 kW) is presented in Section 4.5.1
(Table 4.5.2). Some other estimations for the specific ports, which are further smaller, are
presented in port analysis part (Chapter 5).
The maximum specific energy release in the PFC-windings is well below the limitation
specified in DRG1.
10.1.2 Shutdown Dose Rates outside the Ports
Table 10-2 summarises the maximum dose rates at two weeks after shutdown expected at the
end of the life at locations around the upper, equatorial, and divertor ports where hands-on
maintenance is expected to be performed. The table includes recommendations to improve
dose rate values when necessary.
The dose rates at many locations are low enough (

ITER G 73 DDD 2 01-06-06 W0.1
• NBI port: The very high dose rates are observed around the port near the port extension.
This comes from a week point which locates overlapping part of the port walls from the
vacuum vessel and the ion source.
• Dose rates around the divertor port looks too high suggesting proper shielding
configuration in the port.
Considerable work is still required updating calculation models for following up design
improvement and/or further detailed development of each component design.
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ITER G 73 DDD 2 01-06-06 W0.1
Table 10-2 Shutdown Dose Rates at Maintenance Locations
based on DRG1 Pulse Scenario
Cooling time (~ 10 6 s, first wall neutron fluence 0.3 MWa/m 2 )
Port and Location
1.1 ECH Ports:
Dose Rate
at Two Weeks
After Shutdown
(μSv/h)
1. UPPER PORTS
Comments, Sections
Around the Ports 100 - 800 Improvement of shield is required
minimising void region at the root of the
port (Sec. 5.1)
Primary closure plate TBD Shielding effect of the plug is expected to be
good providing low enough dose rate
2.1 ICRF Ports:
2. EQUATORIAL PORTS
Sec 5.4
At the Closure Plate ~ 200
At the Cryostat
~ 50
Around the Port
30 - 70
2.2 NBI Ports: Sec. 5.2
At the TF Coil Break Boxes ~100
At the Cryostat
~300 Configuration of port extension requires
Around the port extension
2.3 ECH Ports:
200 - 900 improvement.
At the Closure Plate
~130 Sec.5.3
At the Cryostat
~ 30
Around the Ports
2.4 Remote Handling Ports:
70 - 700 Dose at access location is low enough
Around the ports 10 - 20 Sec. 5.7
2.5 TBM ports
Sec.5.6
At the Cryostat
~50
Around the Ports
2.6 Diagnostic Ports,
at the Inner Cryostat Surface:
~100
LIDAR System ~80 Sec.6.3
Motional Stark Effect Diagnostic 60 -80 Sec.6.4
3. DIVERTOR PORTS
3.1 Pumping Ports:
At the Cryostat
3.2 Remote Handling Ports:
~230 Additional shielding around the
regeneration pump. Sec.5.8
At the Cryostat ~60 Sec.5.8
10.2 Conclusions
A fairly sophisticated nuclear analysis has been performed on ITER by means of the most
detailed models and the best assessed nuclear data and codes. This has mainly been focused
on:
• global and local nuclear heating for the component design;
• global and local shielding optimization;
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ITER G 73 DDD 2 01-06-06 W0.1
• radiation conditions in different plasma heating and diagnostic systems;
• radiation conditions in and around the divertor port;
• activation of materials including the cooling water.
In the area of nuclear heating in the superconducting magnet system it has been seen that the
principal source is caused from neutrons and promptly emitted gamma-rays. The total amount
of TFC heating has been computed to be less than 13 kW including the effect of shielding
penetrations such as VV ports. The main contribution to this heating is localised in the inner
leg of the magnet where the shielding has been optimised so to reduce the overall radial build
of the reactor. The volumetric local nuclear heating as well as the radiation damage to the
copper and insulation materials of the magnet has been computed to be far below the
respective limits.
In the light of the sufficiently thick blanket, helium production in the regions of required reweldability
in the vacuum vessel has also been evaluated to be within limits at the end of the
component life time.
Another important area of consideration, in view of its hands-on maintenance consequences,
has been the dose rate for maintenance inside and outside the cryostat shell. In fact neutrons
do activate the reactor components during operation but, as a consequence of the presence of
sufficient shielding, in most of the places the residual dose rate two weeks after shutdown is
on the level of the target for personnel access (100 μSv/h). Some local improvements have
been identified to be required, in particular in the area around the upper port and NBI system,
but no fundamental problems are foreseen.
In summary, the ITER nuclear responses have been evaluated to be sound in all respects
including magnet nuclear heating, radiation damage, activation, vessel helium production,
etc. Further work in this area is needed to verify the local responses of components still to be
developed in full detail, such as port plugs and diagnostics systems.
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ITER G 73 DDD 2 01-06-06 W0.1
Appendix A Chemical Compositions of Materials used for Nuclear
and Occupation Safety Analysis
See Chapter 2 of the report and the main reference 1 .
Additional information on material components may be found in reference 2 .
Table A-1 SS 316L(N)-IG
(for In-Vessel Components and the Vacuum Vessel)
Density: 7.97 x 10 3 kg/m 3
Element wt.% Comments Element wt.% Comments
Basic alloying elements Typical trace elements
Fe balance Al 0.05
C 0.0225 O 0.002
Mn 1.80 K 0.0005
Ni 12.25 Bi 0.0008
Cr 17.50 V 0.004
Mo 2.50 Zr 0.002
N 0.07 Ag 0.0002
Specified impurities Cd 0.0002
P 0.025
S 0.0075 Sn 0.002
Si 0.50 Sb 0.0005
Nb 0.1 - for in-vessel
Ba 0.0005
components
0.01 - for VV only Tb 0.0005
Ta 0.01 W 0.001
Ti 0.15 Ir 0.0005
Cu 0.1 Pb 0.0008
Co 0.05 As 0.0005
B 0.002 - for first wall
0.001 - for VV and cooling
tubes
1 G. Kalinin and V. Barabash, Chemical Composition of Materials for ITER Components. G 73MD 40 00-10-
20 W 0.2. 20 October, 2000.
2 G. Kalinin and V. Barabash, Material Assessment Report. G 74 MA 10 00-11-10 W 0.1.
November 2000.
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Table A-2 SS30467 (304B7)
(the Boarded Shielding Steel for the Vacuum Vessel Filler)
Density: 8.03 x 10 3 kg/m 3
Element wt.% Comments Element wt.% Comments
Basic alloying elements
Not specified impurities
and specified impurities
required for safety analysis
Fe balance O 0.002
C 0.08 K 0.0005
Mn 2.00 Bi 0.0008
P 0.045 V 0.004
S 0.030 Zr 0.002
Si 0.75 Ag 0.0002
Cr 19.0 Cd 0.0002
Ni 13.5 Sn 0.002
N 0.1 Sb 0.0005
B 2.00
Additional requirement
Ba 0.0005
to limit corrosion product activation
Co 0.05
Not specified impurities
required for safety analysis
Nb 0.01 Tb 0.0005
Ta 0.01 W 0.001
Ti 0.15 Ir 0.0005
Cu 0.1 Pb 0.0008
Al 0.05 As 0.0005
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Table A-3 Be S-65 C VHP (Brush Wellman Inc. USA)
(for the First Wall and the Divertor)
Density: 1.82 x 10 3 kg/m 3
Element wppm Comments Element wppm Comments
Major elements, included in
manufacturer specification
Other metallic and non metallic
impurities, guarantied total number less
than 400 wppm
Be Balance W 100
BeO 1.0 U 85
C 0.1 Mo 20
Fe 0.08 Cr 65
Al 0.06 N 225
Si 0.06 Zr 75
Mg 0.06 B 2
Other metallic and non metallic
impurities, guarantied total number less
than 400 wppm
Zn 10 Cd 2
Ni 145 Li 3
Mn 20 Na 15
Sc 5 S 10
Cu 70 F 5
Ag 3 Cl 5
Ti 105 P 0.46
Co 9 Nb 0.12
Pb 20 Ta 9.7
Ca 20 Hf 0.030
Table A-4 Tungsten, pure (manufacturer Plansee AG)
(for the Divertor Targets)
Density: 19.3x10 3 kg/m 3
Element wppm Element wppm Element wppm
W Balance Cu 10 Ni 20
Ag 5 Fe 30 O 30
Al 15 H 5 P 50
As 5 K 10 Pb 10
Ba 10 Mg 5 S 5
C 30 Mn 5 Si 20
Ca 10 Mo 100 Ta 10
Cd 10 N 10 Ti 10
Co 10 Na 10 Zn 5
Cr 10 Nb 10 Zr 10
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Table A-5 Cu Alloys
(for Plasma Facing Structures)
DS Cu alloy, CuAl25-IG PH Cu alloy, CuCrZr-IG
(for the First Wall) (for the Divertor Targets)
Density: 8.86 x 10 3 kg/m 3 Density: 8.92 x 10 3 kg/m 3
Element wt.% Element wt.%
Alloying elements Alloying elements
Cu 99.5 Cu balance
Al (as Al2O3) 0.25 Cr 0.75
O (as Al2O3) 0.22 Zr 0.11
B (as B2O3) 0.025 O 0.03
Specified impurities Specified impurities
Element wppm Element wppm
Class I elements, total < 250 wppm
Pb 10 Al 0.0016
Cd 1 Co 0.06
Zn 5 Fe 0.009
Se 30 P 0.0069
Te 20 Pb 0.0017
P 2 S 0.0023
S 10 Si 0.011
Fe 22 Zn 0.0069
Class II elements, total < 100 wppm
Bi 2
As 10
Sb 10
Sn 9
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Table A-6 Magnet Cable Composition
CS Coil (CS-1) TF Conductor
Cable (316LN density 7.92x10 3 kg/m 3 ,
300K,
Local Volume Fraction) : 0.36 0.36
vol.% vol.%
SC Strand 45 26
Cu wire 7 11
316LN 8 4
He ( 0.1248x10 3 kg/m 3 , 4.2K) 40 24
Incoloy - 13
Insulator - 22
Strands (8.94x10 3 kg/m 3 , 300K) wt % wt %
:
Cu 80 80
Nb 12 12
Sn 4 4
Ta 3 3
Ti 0.2 0.2
Cr 1.0 1.0
Insulator GRP(SC/EP) (1.85x10 3
kg/m 3 , 300K) :
wt % wt %
Glass 57.0 57.0
EP 43.0 43.0
Glass : wt % wt %
Si 37.8 37.8
O 56.7 56.7
B 4 4
Al 1.2 1.2
K 0.3 0.3
EP : wt % wt %
H 3.8 3.8
C 19.6 19.6
N 2.5 2.5
O 37 37
Mg 2.5 2.5
Al 9.0 9.0
Si 19.3 19.3
S 1.4 1.4
Cu 4.9 4.9
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Table A-7 Cryogenic Steels
SS EK1 (European Kind Number 1) SS 316LN
(for the TFC cases, PFC jackets, (for PFC and CS supports,
gravity supports, feeders intercoil
structures, CS
Cryostat)
Density: 8.0 x 10 3 kg/m 3
Density: 8.03 x 10 3 kg/m 3
Element wt % Comments Element wt % Commen
ts
Alloying elements Alloying elements
Fe balance Fe, balance
C 0.03 C 0.03
Mn 1.9 Mn 2.0
Si 0.5 Si 0.75
Cr 17.5 Cr 17.0
Ni 13.75 Ni 12.0
P 0.03 P 0.045
S 0.01 S 0.03
Mo 2.6 Mo 2.5
N 0.22 N 0.14 C+N <
0.22
wt%
Note:
For superconductor jacket, magnet structures, poloidal field coil and divertor the Co and Nb
contents are limited by 0.05 wt.% (500 ppm) and 0.01%, respectively.
Other impurities not included in the specifications are similar to SS 316L(N)-IG for safety
analysis purposes.
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Table A-8 Concrete for the Bio-Shield
- the ANS standard specified for LWR, see reference 1 .
Density: 2.32x10 3 kg/m 3
1 ANS 6.2.1: Shielding Benchmark Problems
Element wt %
H 0.555
O 49.748
Na 1.708
Mg 0.256
Al 4.691
Si 31.471
S 0.128
K 1.922
Ca 8.283
Fe 1.238
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ITER G 73 DDD 2 01-06-06 W0.1
Appendix B Methodologies for dose rate calculation
B.1 Introduction
One of the relevant issues for the ITER machine is the wait time after shutdown until
personnel can access to the cryostat for maintenance. A reliable assessment of the dose rate
calculations is needed, with a suitable procedure.
Two methods have been developed at the purpose: the so-called “two step method”
developed in its full complexity by FZK and Kurchatov Institute indipendently and the “onestep
method” which is quite simpler and faster to be used in actual design calculation. The
latter has been developed in joint collaboration between the JCT and ENEA.
B.2 The two-step method
The shutdown dose rate is due to the decay photons emitted by radioactive nuclides,
generated during the irradiation. In the two step approach the decay nuclides are calculated in
an inventory and activation calculation in which the distribution of the decay gamma sources
are calculated in function of space and time.
The all procedure consists of three steps. A first step is necessary in which neutron transport
calculation is performed by MCNP to get the spatial distribution of the neutron flux,
generally in 175 groups. Then in a second step an inventory calculation is performed for the
radioactive nuclide inventory and the decay gamma source using the neutron flux obtained in
the first step. The second step is performed by means of FISPACT 1 code with FENDL-2.0/A 2
activation library. The decay gammas are calculated for all the cells of the system, even if in
a group structure (24 or 22 groups). In the third step MCNP is used again to transport the
gamma, making use of the spatial decay gamma source distribution.
The procedure is straightforward, in practice in large complex system like ITER there are a
thousand of cell or more. The most difficult part is to define in a correct way all the gamma
decay source with the proper normalization. This is impossible to be achieved manually.
Interface programs have been developed by FZK to link MCNP and FISPACT and vice versa
to solve the problem. More can be found in 3 , 4 .
The main advantage of the two step method is that all the possible radioisotopes contributing
to the dose rate are considered with no theoretical restrictions. Multi-step reactions are
considered. A main drawback, is the approximate assumption of gamma source spatial
distribution, which may give significant effect in large geometry calculation. Other possible
drawback is multigroup approach, especially in the decay gammas. It is also not clear how to
calculate the statistical error of the resultant dose rate. The simple addition of the fractional
standard deviation of the two MCNP calculations does not look correct.
1 RA Forrest, J-Ch Sublet: FISPACT 99: User Manual UKAEA Fusion, Report UKAEA FUS 407, Dec 1999.
2 A.B. Pashchenko, H. Wienke, J. Kopecky, J-Ch Sublet and R.A. Forrest,: FENDL/A-2.0 Neutron Activation
Cross Section Data for Fusion Application IAEA report IAEA-NDS-173, Rev. 1 Oct 1998.
3 Y. Chen, U. Fisher ITER-FEAT Shutdown Dose Rate Analysis by Rigorous Method. Draft Final Report on
Contracty FU05-CT 2000-00134 June 2001
4 G. E. Schatalov, A.A. Borisov, I. A. Kartashev, A. G. Serikov, S. V. Sheldyakov, O. L. Schipakin, Neutronic
Analysis of the ITER Vacuum Vessel/ Cryostat Environment, Quarter reports JF –04-00/1, April 2000, and JF
04-01/1 March 2001
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ITER G 73 DDD 2 01-06-06 W0.1
B.3 The one-step method
This method was proposed to have exact spatial distribution of decay gamma source.
According to this new methodology the gamma rays emitted by the radioactive isotopes are
coupled to the neutrons as they were promptly emitted. At the purpose a special format crosssection
library is needed and a modified version of MCNP. The neutron part is like a normal
transport library, but as regard the photon generation part, only the cross sections generating
radioisotopes are included. Absolute and relative intensities, energy spectra of the gamma are
taken according to the decay channel. This trick allows the decay gamma rays to be
transported, as they were prompt. The eventual energy release is scored.
An artificial delay time, expressed in shakes (1 shake = 10 -8 sec), is given to each different
gamma to distinguish the different gamma rays coming from the different isotopes. A proper
time binning of the gamma related tallies in MCNP shares the contributions of the different
isotopes. The time delay is artificial, it has nothing to do with the decay time of the
radioactive isotope; the purpose is only to distinguish the different gamma ray origins.
For dose rate calculations the user has to couple the contributions of the different isotopes by
a factor, which is function essentially of the time. This factor takes into account the build up
of the isotope during irradiation and its decay at shut down. It can be derived in an analytical
way or derived more exactly using properly the outcome of the inventory codes. THIDA-2
1 or FISPACT have been used so far. More details can be found in 23 4 and 5 .
The single step approach has the advantage to divide the time dependency of the radioactive
decay from its spatial dependency. Moreover there is a real coupling of the neutron transport
with the emitted gamma. This is reflected in the statistical error, which in the one-step direct
method is really the outcome of all the processes involved.
The new proposed method, at the moment, has some limits. For example multi-step isotopes
formation or calculations in which there is a strong high burn-up cannot be handled.
Moreover, the cross section library needed for dose rate calculation has at the moment a
reduced number of isotopes. This is sufficient for the dose rate calculation 10 6 sec after
shutdown in ITER, when most of the activation comes from several isotopes formed in
stainless steel. For a different elapsed time and a different material composition the library
should be improved for more general application.
1 Y. Seki, H. Iida, H. Kawasaki, K. Yamada: “THIDA-2: An advanced Code System for Transmutation,
Activation, Decay Heat and Dose Rate”, Japan Atomic Energy Research Institute, JAERI 1301, March 1986.
2 H.Iida, D. Valenza, R. Plenteda, R. T. Santoro and J. Dietz "Radiation Shielding for ITER
to allow for Hands-on Maintenance inside the Cryostat", J. of Nuclear Science and
Technology, Supplement 1. p.235-242 (March 2000)
3 D. Valenza, H. Iida, R. Plenteda: "Proposal of Shutdown Dose Estimation Method by Monte Carlo Code" to
be published in the journal of Fusion Engineering and Design
4 L. Petrizzi, H. Iida, D. Valenza, and P. Batistoni: Improvement and benchmarking of the new shutdown dose
estimation method by Monte Carlo code. Presented at MC2000 conference Advanced Monte Carlo for
Radiation Physics Particle Transport Simulation and Applications. 23-26 Oct. 2000 Lisbon, Portugal--
5 L. Petrizzi, H. Iida, D. Valenza: “ Further development of a method for calculating the dose rate by means of
MCNP” NAG-143-13-12-99
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ITER G 73 DDD 2 01-06-06 W0.1
Appendix C ITER FEAT MCNP Models
C.1 Introduction
During the 6 years of ITER EDA (1992-1998), a full 3-D model has been set-up with full
details of any component. The basic model had 9˚ symmetry because the machine was made
of 20 TF coils.
Then the objective was to study a smaller machine with a lower fusion power. At that point
more parametric one-dimensional calculations were needed to understand more or less the
thickness needed to shield in a proper way the outer machine. Rough 3 D models were built
up just to have scaling laws of intermediate studies like IAM or LAM. In 2000 the work
started for the construction of a new 3 D model of an assessed design: ITER FEAT. In a short
period a high degree of complexity has been reached thanks to the shared work between the
Home Teams. Each has developed an own part and the overall has been assembled by the
JCT.
C.2 ITER 20˚ basic model
The model has been built in two phases. The first has been completed in June 2000, a more
complete report is in the memo 1 . The second phase ended by the end of the year 2000 with
the completion of the outer part of the machine (e.g. the cryostat) 2 for a total of more than
2800 cells!.
The basic model consists of the minimum symmetry portion of the reactor that can be used
for estimating nuclear responses in 3-D calculations that is 20˚ (1/18 of the reactor, as there
are 18 TFCs). A picture of the model, according to the first stage is in Figure C-1.
The following components have been described:
• Blanket system: 45 cm thick modules consisting of a First Wall (FW: 1 cm thick Be
armour and 2 cm thick plate of Cu, water and SS) and Bulk Shield (BS: 42 cm SS and
water). Toroidally and poloidally 2 cm wide gaps are running between the modules; there
are poloidal filling wedges between some modules.
• Blanket Support: inboard and outboard blanket supports in the lower part close to the
divertor;
• The Vacuum Vessel: a three zones structure consisting of 6 cm SS inner shell, borated SS
and cold water filler zone and 6 cm SS outer shell ;
• Upper Ports: two half ports;
• Equatorial Ports: two half ports, in its preliminary configuration (no antennae or other
devices);
• Divertor Ports: two half ports different in shape, one for the remote handling the second
for the cryo-pump.
• Divertor: 2 whole cassettes and 2 half cassettes (3 cassettes in 20 degrees), the gap
between the cassettes is 1 cm wide. The divertor system has been modeled by L.Petrizzi
(EU HT), according to the 2000 model, with high level of detail.
1 G. Ruvutuso and H. Iida, NAG-159-08-06-00, “Three-Dimensional model of the ITER-FEAT reactor for
Monte Carlo nuclear analyses with MCNP”.
2 G. Ruvutuso, L. Petrizzi and H. Iida NAG-168-14-11-00: “Updated Basic 3-D model of ITER for Monte Carlo
nuclear analyses with MCNP”
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ITER G 73 DDD 2 01-06-06 W0.1
• A whole Toroidal Field Coil (TFC): there is one whole TFC in the 20˚ model. The TFC
winding pack and filling materials are homogenised. The insulator layer has been
described. The TFC case is stainless steel. The SS intercoil structures located between the
ports were added.
• The Central Solenoid (CS) and 6 Poloidal Field Coils (PFC).
Figure C-1 Poloidal section of the 3D ITER-FEAT model through the gap in
between the divertor cassettes
In the second phase the basic model has been revised and extended. The main modifications
were relative to
3. the source cell, switching from the 5 cells source definition with uniform sampling inside
each of them to the external point-wise source given as an external subroutine source;
4. updated toroidal segmentation of blanket modules (figure C-2), with detailed description
of manifolds running behind the modules, coming in the in-vessel components through
the upper port. The blanket segmentation is such that there are 17 modules in poloidal
direction. From modules 1 to 8 (inboard) there are 2 half modules in the 20˚ model, to
keep the toroidal position respect to the ports. In the outboard (from 9 to 17) there are one
whole and two halves. 20-mm-wide poloidal and toroidal gaps separate all of the
modules.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure C-2 3D views by SABRINA of blanket modules and filler system.
5. modifications of the upper port shape;
6. update of PFC ;
7. addition of the port extensions: upper ports, equatorial ports and lower ports (RH port &
Cryopump port) and cryostat, the cryo-pump system itself. There are six half ports in the
model: two at the divertor level, two at the equatorial mid-plane and two at the upper
region. The Neutral Beam Injector (NBI) ports are not included since they are not normal
to the plasma major axis. A separate model has been developed for the NBI studies.
8. new gravity support.
The resulting model can be seen in Figures C-3. The geometry of the divertor is not
consistent with the latest design, it will be updated as soon as the related design activity will
end.
C.3 Neutron Source definition
A Gaussian distributed D-T neutron source, peaked around 14.1 MeV, was used. The neutron
source intensity is sampled, in the latest model version, according to an external FORTRAN
subroutine and depends on the position. The intensity information were collected in a 40x40
R-z array.
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ITER G 73 DDD 2 01-06-06 W0.1
Figure C-2 Poloidal cross sections of the 3D model through the symmetry
planes of the remote handling port and the cryostat port
C.4 Material compositions
The following tables give the compositions of the most used basic materials in the ITER
model. They are consistent with the chemical composition given in Appendix A.
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ITER G 73 DDD 2 01-06-06 W0.1
Table C-1 a,b : Stainless Steels compositions (Unit: atoms/cm 3 ) in 10 24
S.S.316-ln S.S.316-lw
t.a.d.= 8.67517E-02 t.a.d.= 8.5329E-02
B10 1.517E-06 B10 8.7072E-07
B11 5.552E-06 B11 3.5047E-06
C 7.148E-05 C 3.9383E-05
N14 1.907E-04 N14 6.4166E-04
O16 4.771E-06 O16 5.8990E-06
Al27 7.071E-05 Al27 5.2595E-04
Si 6.793E-04 Si 1.6843E-04
P31 3.080E-05 S 2.0654E-06
S 8.925E-06 K 6.0493E-07
K 4.879E-07 Ti 3.9519E-05
Ti 1.195E-04 V 3.7143E-06
V 2.996E-06 Cr50 6.7323E-04
Cr50 5.762E-04 Cr52 1.3084E-02
Cr52 1.077E-02 Cr53 1.4917E-03
Cr53 1.205E-03 Cr54 3.7176E-04
Cr54 2.946E-04 Mn55 1.2743E-03
Mn55 1.250E-03 Fe54 3.2814E-03
Fe54 2.675E-03 Fe56 5.1513E-02
Fe56 4.049E-02 Fe57 1.2193E-03
Fe57 9.416E-04 Fe58 1.7418E-04
Fe58 1.322E-04 Ni58 6.6297E-03
C059 3.237E-05 Ni60 2.5595E-03
Ni58 5.467E-03 Ni61 1.2230E-04
Ni60 2.040E-03 Ni62 3.5810E-04
Ni61 9.589E-05 Ni64 1.1350E-04
Ni62 2.762E-04 Zr90 5.3368E-07
Ni64 8.481E-05 Zr91 1.1646E-07
Cu63 1.257E-04 Zr92 1.7744E-07
Cu65 5.448E-05 Zr94 1.8045E-07
Zr90 4.368E-07 Zr96 2.9038E-08
Zr91 9.427E-08 Nb93 2.8003E-05
Zr92 1.421E-07 Mo 1.0010E-03
Zr94 1.414E-07 Sn 7.9695E-07
Zr96 2.228E-08 Ta181 1.3071E-07
Nb93 4.107E-06 W182 6.7844E-08
Mo 9.943E-04 W183 3.7006E-08
Sn 6.428E-07 W184 7.8637E-08
Ta181 2.109E-07 W186 7.2984E-08
W182 5.536E-08 Pb206 4.1108E-08
W183 3.003E-08 Pb207 4.1273E-08
W184 6.347E-08 Pb208 9.7411E-08
W186 5.827E-08
Pb206 3.336E-08
Pb207 3.333E-08
Pb208 7.829E-08
Bi209 1.461E-07
Nuclear Analysis Report Page 218

ITER G 73 DDD 2 01-06-06 W0.1
Table C-2 : a,b compositions (Unit: atoms/cm 3 ) in 10 24 of VV filler (borated steel 60% and
water 40%), concrete
a) b)
SS Borated Stee 60%
+40% water
Concrete
At density= 9.28e-02 T density =7.38e-2
B10 1.140E-03 H1 7.700E-03
B11 4.160E-03 H2 1.200E-06
C 1.910E-04 O16 4.300E-02
N14 2.040E-04 Na 1.000E-03
O16 3.580E-06 Mg 1.500E-04
Al27 5.300E-05 Al 2.400E-03
Si 7.640E-04 Si 1.570E-02
P31 4.620E-05 S 5.600E-05
S 2.680E-05 K 6.900E-04
K 3.660E-07 Ca 2.900E-03
Ti 2.990E-05 Fe54 1.230E-05
V 2.250E-06 Fe56 1.930E-04
Cr50 4.690E-04 Fe57 4.560E-06
Cr52 8.770E-03 Fe58 6.510E-07
Cr53 9.810E-04
Cr54 2.400E-04
Mn55 1.040E-03
Fe54 1.930E-03
Fe56 2.920E-02
Fe57 6.780E-04
Fe58 9.520E-05
Co59 2.430E-05
Ni58 4.520E-03
Ni60 1.690E-03
Ni61 7.930E-05
Ni62 2.280E-04
Ni64 7.010E-05
Cu63 3.140E-05
Cu65 1.360E-05
H1 2.670E-02
H2 5.350E-06
O16 1.340E-02
Nuclear Analysis Report Page 219

ITER G 73 DDD 2 01-06-06 W0.1
Table C-3 : Composition (Unit: atoms/cm 3 ) in 10 24 of Magnetic conductor (homogenised)
Magnetic conductor
At density= 7.1943eE-2
H1 3.890E-03
C 3.410E-03
N14 3.710E-04
O16 4.870E-03
Mg 2.140E-04
Al27 7.070E-04
Si 1.440E-03
S 9.180E-05
Cu63 1.130E-04
Cu65 5.060E-05
Cu63 6.840E-03
Cu65 3.060E-03
Nb93 1.180E-03
Sn 3.950E-04
B10 4.060E-07
B11 1.490E-06
C 1.700E-04
N14 2.770E-04
O16 2.560E-06
Al27 2.270E-04
Si 7.280E-05
P31 1.650E-05
S 8.930E-07
K 2.610E-07
Ti 1.710E-05
V 1.610E-06
Cr50 3.030E-04
Cr52 5.670E-03
Cr53 6.340E-04
Cr54 1.550E-04
Mn55 5.580E-04
Fe54 1.470E-03
Fe56 2.220E-02
Fe57 5.160E-04
Fe58 7.240E-05
Co59 1.730E-05
Ni58 2.890E-03
Ni60 1.080E-03
Ni61 5.070E-05
Ni62 1.460E-04
Ni64 4.490E-05
Cu63 2.240E-05
Cu65 9.730E-06
Zr90 2.340E-07
Zr91 5.050E-08
Zr92 7.610E-08
Nuclear Analysis Report Page 220

ITER G 73 DDD 2 01-06-06 W0.1
Zr94 7.580E-08
Zr96 1.190E-08
Nb93 1.210E-05
Mo 4.260E-04
Sn 3.440E-07
Ta181 5.650E-08
W182 2.970E-08
W183 1.610E-08
W184 3.400E-08
W186 3.120E-08
Pb206 1.790E-08
Pb207 1.790E-08
Pb208 4.190E-08
Bi209 7.830E-08
Cu63 3.390E-03
Cu65 1.520E-03
Sn 2.630E-04
Nuclear Analysis Report Page 221