Methodology to address radioprotection and safety issues in

the IFMIF/EVEDA accelerator prototype

 

J. Sanz1,2, P. Sauvan1,2 F. Ogando1,2, D. López1,2, M. García1,2, A. Mayoral1,2, F. Ortiz1

A. Ibarra3, V. Blideanu4, P. Joyer4

(1)Departamento de Ingeniería Energética, Escuela Técnica Superior de Ingenieros Industriales,

Universidad Nacional de Educación a Distancia (UNED), C/ Juan del Rosal 12, 28040 Madrid

(2) Instituto de Fusión nuclear, Universidad Politécnica de Madrid (UPM), C/ José Gutiérrez Abascal 2, 28006 (Madrid)

(3) CIEMAT, Madrid, España

(4) Commissariat à l’Energie Atomique, CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex, France

Shielding Aspects of Accelerators, Targets and Irradiation Facilities (SATIF-10) Geneva, Switzerland, June 2–4, 2010,

•Radioprotection and safety issues in the IFMIF/EVEDA accelerator prototype

• General Computational Methodology Flowchart: issues to discuss

•Testing available deuteron cross sections: Benchmarking against experimental EXFOR data * Validation of nuclear models included in MC transport codes: MCNPX,PHITS * Interest of the TALYS nuclear model code to simulate deuteron reactions

• MCUNED: new capabilities for Monte Carlo simulation of light ions transport and secondary products generation

* New capabilities: extensions to MCNPX* Verification of MCUNED capabilities

•MCUNED applications to the validation process of TENDL for deuterons: Analysis of integral experiments for neutron production

•Proposals under discussion to deal with other RP issues

• Summary

Overview of presentation

IFMIF/EVEDA prototype accelerator(Project under Fusion Broader Approach Japan/EU)

IFMIF: International Fusion Materials Irradiation Facility2 deuteron beams (125 mA X 2) up to 40 MeV

EVEDA phase: Engineering Validation and Engineering Design Activities1 deuteron beam (125 mA) up to 9 MeV

BD

ECRsource

LEBT RFQ

beam

DTL

~ 10 m ~ 5 m

ECRsource

LEBT RFQ

beam

DTL

~ 10 m ~ 5 m

100 keV165 mA 5 MeV

140 A9 MeV125 mA

Faraday Cupnormal incidence

Collimator in normal operationor interceptive diag at low d.c. (10-3)

normal incidence

Beam Dumpangle of incidence

according to design

Loss in RFQ~ 10 mA max DTL and HEBT

1 W/m ~ 100 nA/m

Potential radioprotection issues for the IFMIF/EVEDA prototype accelerator

Some radioprotection issuesBeam-on phase:* Generation of neutrons and gammas by d-target material and d-D interactions: dose levels inside and outside the vault* Tritium production Beam-off Phase:* Deuteron and neutron induced activation: Residual gamma dose rates

BD cartridge: Definition of local shielding M. García Talk

Concrete walls of the accelerator vault: Thickness = 1.5 meters

CONCRETE SHIELD

COMPOSITION

0.56 % H

2.21 g/cm3

General Computational Methodology: calculational tools

Particle TRANSPORT (MC codes)

Activation System

Neutron and deuteron fluxes

Gamma TRANSPORT

Isotopic inventoryGamma Source

Gamma Dose

Neutron Source

d-D

Neutron Sourced-Cu

Beam dynamics code

Module for neutronSource modeling

Module for neutron Source modelingModule for

Deuteron source modeling

BE

AM

O

FF

Neutron and gamma Fluxes

Neutron and gammaDoses

Dose conversion factorsB

EA

M

ON

Deuteron transport, interaction and secondary particle production

* Monte Carlo transport codes: built-in semi analytical models to describe nuclear interactions of deuterons with mater.

- For incident deuterons of energy relevant to IFMIF/EVEDA accelerator:

* are the models reliable enough?* can computing time consumption be a concern?

Nuclear reaction codes applications to calculate deuteron cros sections

• Explore de interest for generation of evaluated XS libraries by dedicated nuclear model codes: TALYS code and TENDL library

Validation of deuteron nuclear dataValidation of deuteron nuclear data

• Validation (first step: basic experimentally measured nuclear data; EXFOR)• Validation (second step: integral experimentally measured data)

Reliability of existing methodologies: Availability, development a validation of deuteron nuclear data

MC transport code LIMITATIONS: MCNPX and others (PHITS)

• Elements studied : Typical of accelerator equipments Cu, Ni, Fe, Cr, Nb and W • ENERGY RANGE : 0 to 20 MeV

NO nuclear models in MCNPX ALLOW A GOOD FIT WITH EXPERIMENTAL VALUES (in shape and amplitude). Strong disagreement is observed for main reactions

Relevant impact on the prediction of the secondary neutron source production

In a few cases MCNPX cross sections are in reasonable agreement with experiments, but with an unreliable spectrum of emitted neutrons

MCNPX generate unphysical high energy neutron tails above the maximum physical energy (about 15.5 MeV for Cu63)

Big impact on neutron activation (production of radionuclides of concern from neutron threshold reactions)

MC transport code LIMITATIONS: MCNPX and others (PHITS)

Spectrum of the emitted neutrons is unphysical: high energy tail is computed

MC CODES LIMITATIONS: NEUTRON EMISSION

Natural Cu

WHAT CAN THE TALYS CODE (and TENDL) OFFER?

TALYS (global OM parameters) gives a shape that fits rather well the shape of experimental values TENDL=TALYS based library

However, the results are in general not good enough in amplitude (not much better than in MCNPX in several cases)

Potential to change Optical Model parameter in order to explore the possibility to obtain a right fit to experimental data

Even in those cases in which TALYS results are not good enough in amplitude, neutron spectrum is consistent with energy conservation

POTENTIAL OF TALYS CODE (and TENDL)

Spectrum of the emitted neutrons is consistent with energy conservation

TALYS-TENDL: NEUTRON EMISSION

Natural Cu

USE OF APPROPRIATE ADJUSTING PARAMETERS CAN ALLOW A GOOD REPRODUCTION OF EXPERIMENTAL CROSS SECTIONS?

The fitting set of parameters has to be defined for each element.

Figures show the improvement obtainedwith the fitting set of parameter:

56Fe (UNED) 65Cu (Avrigeanu)

One of the best know transport code is MCNPX

They have some problems for IFMIF-EVEDA use:

1) NO nuclear models in MCNPX ALLOW prediction of excitation functions showing A GOOD FIT WITH EXPERIMENTAL VALUES for interactions of D with typical elements of accelerator equipments in the energy range of interest

2) MCNPX provide unreliable spectrum of emitted neutrons * unphysical high energy neutron tails above the maximum physical energy

* for some models not able to provide neutrons for deuteron incident energies below 5-6 MeV

3) Computing time consumption concerns for the low neutron yields in the EVEDA scenario

Similar problems can be found in other standard transport codes (for example PHITS or FLUKA)

Why extensions to current MC transport codes ?

External light ions evaluated libraries (for example TENDL –generated from TALYS code-) must be used

We need a code able to read external libraries

McUNED

MCUNED features and capabilities

MCUNED is a MCNPX extension allowing handle light ions evaluated libraries and incorporating a variance reduction technique for the production of secondary particles by light ions induced nuclear reactions

• Able to handle proton, deuteron, triton, helium-3, alpha evaluated nuclear data libraries for nuclear reactions in transport calculations.

• Include a variance reduction technique for the production of secondary particles. Save drastically the computing time.

• Maintain all MCNPX capabilities :– Same input as MCNPX (fully compatible with MCNPX input) – Same particle transport algorithm– Same flexibility in source and geometry definition– Same calculation capabilities (Flux, dose, mesh etc… tallies)

Verification of handling light ion libraries

9 MeV deuteron

Cu63 target

0.1 m thickness

2 4 6 8 10

10-3

10-2

Ed=9MeV

=0º

Pro

babi

lity

(1/M

eV/S

r)

2 4 6 8 10

10-3

10-2

=30º

MCUNED Janis 3.0

0 2 4 6 8 10

10-3

10-2

=60º

Pro

babi

lity

(1/M

eV/S

r)

Neutron Energy (MeV)

0 2 4 6 8 10

10-3

10-2

=120º

Neutron Energy (MeV)

• Thin Target experiment on Copper 63 Neutron spectrum produced by the simulation are recorded at 4 different angles They are compared with double differential XS data used in the calculations (XS data from Tendl 2008 are extracted with Janis 3.0 converted into laboratory system)

• Satisfactory results

Verification of Variance Reduction Technique

The verification of the Variance Reduction technique is carried out by comparing MCNPX and MCUNED results with incident proton simulations.

Incident proton simulation are used since MCNPX could use only proton data library

The same input file is used in both simulation

The number of initial proton histories in both simulations is adjusted to achieved the same statistical error in the tally value.

The Variance Reduction technique is considered correct if the relative error between MCNPX and MCUNED calculation is comparable to the statistical error of the tally value

MCUNEDMCNPXMCNPX

NEDMCUMCNPX

r axmX

XX ,

Relative error

Statistical error

The same simulation with proton on thick target have been performed by MCNPX and MCUNED 10 MeV

proton

Natural Copper target

1 mm thickness

0 1 2 3 4 5 6 7 8

1E-8

1E-7

1E-6

1E-5

1E-4

Total ( =4)

ne

utr

on

/ p

rotr

on

/ M

eV

Neutron Energy(MeV)

MCNPX MCUNED

2 4 6 8

1E-8

1E-7

1E-6

1E-5

=0º

neut

ron/

pro

ton/

sr/

MeV

2 4 6 8

1E-8

1E-7

1E-6

1E-5

=30º

MCNPX MCUNED

0 2 4 6 8

1E-8

1E-7

1E-6

1E-5

=60º

neut

ron/

pro

ton/

sr/

MeV

Neutron Energy (MeV)

0 2 4 6 8

1E-8

1E-7

1E-6

1E-5

=120º

Neutron Energy (MeV)

The proton history number have been chosen in both cases to reach the same statistical error (about 1%)

The agreement between both simulations is very good

Verification of Variance Reduction Technique

Verification in Response function calculations

Neutron ambient equivalent dose calculated for a 100 mA proton beam @6.7 MeV on cylindrical thick Nickel Target

MCNPX simulationNeutron dose map

MCNPX and MCUNED simulations are performed for dose calculation

• Both results have the same statistical error

MCNPX

NEDMCUMCNPX

r D

DD

MCNPX SimulationStatistical error map

MCNPX ,MCUNED Relative error map

Relative errors between MCNPX and MCUNED mean values are lower than the statistical errors of the given response function

Spatially averaged statistical error is about 4%

Verification in Response function calculations

Performance of the Variance Reduction

The reduction variance technique is useful when the production of secondary particle is low (it is essential when the production rate of secondary particle is very low)

p

s

p

s

T

T

T

T

G1

1 • Number of secondary particle per primary particle• Tp Time needed to transport the primary particle• Ts Time needed to transport the secondary particle

Summary of 10 MeV Proton on Copper simulation

In this case the time gain is 4000 !!!

TENDL Library

TENDL : Talys-based Evaluated Nuclear Data Library

• Library generated by talys nuclear code

• Available libraries for incident neutron, proton, deuteron, triton, helion, alpha

• Incident energy up to 200 MeV

• Data available for 1000 isotopes

•Libraries available in ENDF and ACE format

http://www.talys.eu/tendl-2009/

Checking TENDL library against integral experiments

[1] L. W. Smith and P. G. Kruger, “Thick Target Yields from (d,n) Reaction at 10 MeV”, Physical Review, 83, 1137, (1951).[2] A. J. Allen, etal., “Thick Target Fast Neutrons Yield from 15-MeV Deuteron and 30-MeV Alpha-Bombardement ”, Physical Review, 81, 536, (1951).[3] K. Shin etal., “Neutron and photon production fron thick targets bombarded by 30-MeV p, 33-MeV d, 65-MeV 3He, and 65-MeV a ions: Experiment and comparison with cascade Monte Carlo calculations”, Physical Review C, 29, 1307, (1984).[4] J. P. Meulders, etal. , “Fast Neutron Yields and Spectra from Targets of Varying Atomic Number Bombarded with Deuterons from 16 to 50 MeV”, Phys. Med. Biol., 2, 235, (1975).

Total neutron Yield (4)

RefEd

(MeV)En >

(MeV) Exp(n/d)

TENDL(n/d)

Yield frac.

1 10 0 8.81E-4 6.44E-4 1

2 33 0 1.81E-2 1.50E-2 1

RefEd

(MeV)En >

(MeV) Exp

(n/d/Sr)TENDL(n/d/Sr)

Yield frac.

3 16 4 2.76E-4 5.94E-5 0.20

2 33 4 6.16E-3 7.49E-4 0.39

3 33 4 6.68E-3 7.49E-4 0.39

Neutron Yield in forward direction

Total yields from simulations are in reasonable agreement with experimental values

TENDL underestimates neutrons emitted in the forward direction

Deuteron induced neutron production in a thick copper target

5 10 151E-7

1E-6

1E-5

1E-4 10 20 30

1E-6

1E-5

1E-4

Ed=16 MeV

n/d

/ S

r /

Me

V

En (MeV)

Ed=33 MeV

n/d

/ S

r /

Me

V

MCUNED+TENDL Meulders et al.

Spectra of neutrons emitted in the forward direction

Checking TENDL library against integral experiments

Normalized neutron angular distribution

Thick copper target bombarded by deuterons

• Forward neutron spectra are not reproduced by TENDL

• Experimental angular distribution exhibit a forward peaked neutron emission not reproduced by TENDL .

•Differences become lower for lower incident energies

• Backward neutron spectra from simulation are in good agreement with experiments

0 20 40 60 80 100 120 140

0,05

0,10

0,15

0,00

0,25

0,50

0,75

MCUNED+TENDL Exp. (Allen etal.)

Ed=15 MeV

En > 1 MeV1/

sr

Angle (Deg)

Ed=33 MeV

En > 4 MeV

1/sr

MCUNED+TENDL Exp. (Shin etal.)

0 10 20 30

1E-6

1E-4

1E-6

1E-4

1E-6

1E-4

1E-6

1E-4

1E-6

1E-4

=135º

En(MeV)

=75º

=45º

n/d/

MeV

/sr

=15º

=0º

MCUNED-Tendl09 K.Shin etal.

Checking TENDL library against integral experiments

Thick target experiment with deuteron on Aluminum

105

106

107

108

109

0 10 20 30 40105

106

107

108

109

0 10 20 30 40 50

=0º

neut

rons

/MeV

/sr/C

=30º

MCUNED-Tendl 2008 Exp. M.Hagiwara etal. (2004)

=60º

neut

rons

/MeV

/sr/C

En(MeV)

=110º

En(MeV)

Reasonably agreement with Tendl at low deuteron energy

• T.N.Massey etal. “A Measurement of the 27Al(d,n) Spectrum for Use in Neutron Detector Calibration” Nuclear Science and Engineering, Vol.129, p.175 (1998)• M.Hagiwara etal. “Experimental studies on the neutron emission spectrum and activation cross-section for 40 MeV deuterons in IFMIF accelerator structural elements” Journal of Nuclear Materials, Vol.329, p.218 (2004)

Aluminum 6mm thick, Ed=40 MeVThick Aluminum foil, Ed=7.44 MeV

0 5 100

2x107

4x107

6x107

8x107

=120º

neut

rons

/MeV

/sr/C

En(MeV)

MCUNED - Tendl exp. T.N.Massey etal.- (1998) MCNPX (ISABEL)

Issue: Deuterium Concentration Profile inside the Copper Lattice and n production

Some Proposals under discussion

Deuterium implantation profile: MCNPX/MCUNED, SRIM

Deuteron concentration profile: New transport coefficient data for IFMIF/EVEDA operation conditions and TMAP code calculations

Neutron production: MCUNED.

OTHER ELEMENTS OF THE COMPUTATIONAL METHODOLOGY (still under discussion within the Project)

Issue: Deuteron, proton a neutron induced activation

Some Proposals under discussion

ACAB activation code. EAF data libraries

Issue: Tritium production by deuteron induced reactions (d-D, d-Cu)

Some Proposals under discussion

Tritium production and implantation profile: MCUNED

Tritium concentration in structural materials, and diffusion to water, vacuum pump: Transport coefficient data for IFMIF/EVEDA operation conditions and TMAP calculations.

Summary

Different alternatives have been found regarding the different computational elements needed in addressing the RP issues of the EVEDA accelerator prototype.

Some alternatives are under discussion within ASG between CEA and UNED regarding activation and deuterium implantation issues. Final agreement will be reached soon

Regarding simulation of deuteron transport and secondary products generation the agreement is already reached

1. The nuclear reaction models included in the transport codes (MCNPX, PHITS) cannot be used in the EVEDA RP calculations.

2. The cross-sections of the reactions should be re-assessed with the TALYS code which allows a better description of the nuclear reactions of interest for EVEDA

3. Extension of Monte Carlo codes to use evaluated data files generated by TALYS with appropriate adjusting OMP parameters

4. MCUNED code, a MCNPX extension is the current response to this need.

Summary

MCUNED: Two extensions to MCNPX

* Handling light ion data libraries (energy-angle distributions of all outgoing particles): enables to include updated/reliable nuclear cross section for transport simulation * Reduction variance technique for generation of secondary products: drastic reduction in the computing time needed for a target accuracy

MCUNED has been verified with very positive results

Valuable in radioprotection studies on accelerator design: it is specially targeted for low-energy light ion applications

Superior performance of MCUNED-TENDL libraries versus current Monte Carlo simulation tools for deuterons in the EVEDA energy range: adjusting D-TENDL library required?

Useful tool for Benchmarking evaluated cross section against integral experiment

Summary

Efforts are underway to evaluate TENDL for incident deuterons on copper against integral experiment for neutron production.

Total neutron yields from simulations are in a reasonable agreement with experiments

Angular distribution and spectrum of backward-emitted neutron is well reproduced by TENDL

TENDL underestimates neutrons emitted in the forward direction

Neutron spectra not reproduced for neutrons emitted in the forward direction

Differences between experiments and simulations decreases as incident deuteron energy decreases, and a reasonable behavior of TENDL library could be expected for EVEDA applications

Few experiments exist for EVEDA applications.

Very recent experiments will be very useful in evaluating and improving (if necessary) TENDL for EVEDA applications.