neutronics and nuclear data for the ifmif neutron source

Neutronics and nuclear data for the IFMIF neutron source

U. Fischer a,�, S.P. Simakov a, A. Konobeyev b, P. Pereslavtsev b, P. Wilson c

a Association FZK-Euratom, Forschungszentrum Karlsruhe, Institut fur Reaktorsicherheit, P.O. Box 3640, 76021 Karlsruhe, Germanyb Institute of Nuclear Power Engineering, Obninsk, Kaluga Region, Russian Federation

c Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI 53706, USA

Abstract

An overview is presented of the R&D work conducted at Forschungszentrum Karlsruhe in co-operation with the

Institute of Nuclear Power Engineering, Obninsk, and the University of Wisconsin (UW), Madison, on the

development of neutronic computational tools and nuclear data for the International Fusion Material Irradiation

Facility intense D�/Li neutron source. The focus is on the progress achieved recently for the D�/Li neutron source term

modelling with an advanced Monte Carlo procedure making use of newly evaluated double-differential data for the6,7Li(d, xn) reactions, and the creation of the Intermediate Energy Activation File IEAF-2001, an activation data

library comprising 679 target nuclides from Z�/1 (hydrogen) to 84 (polonium) with neutron induced activation

reactions up to 150 MeV.

# 2002 Elsevier Science B.V. All rights reserved.

Keywords: Fusion; Neutron source; d�/Li reaction; Activation; Nuclear data

1. Introduction

The International Fusion Material Irradiation

Facility (IFMIF) project [1] aims at providing an

intense neutron source for high fluence test

irradiations of fusion reactor candidate materials.

IFMIF employs two continuous-wave linear accel-

erators each generating a 125 mA beam of 40 MeV

deuterons striking a thick target of flowing liquid

lithium to produce high energy neutrons for the

irradiation of material samples at a radiation load

as anticipated for a future fusion power reactor.

Neutronics and nuclear data play a key role in

establishing IFMIF as an intense neutron source

for the development and qualification of fusion

reactor materials: (1) IFMIF’s suitability as neu-

tron source for fusion-specific simulation irradia-

tions must be proven by means of neutronic

calculations, and (2) the technical layout of the

test modules, facility sub-systems, shielding, etc.

relies on the data provided by nuclear design

calculations. This includes the proof that IFMIF

can meet it’s design goal with regard to the

required irradiation test volume as well as the

attainable annual fluence accumulation.

Much effort has been spent over the past few

years to develop the computational tools and

nuclear data required to enable proper neutronic

calculations for the IFMIF neutron source. There

� Corresponding author. Tel.: �/49-7247-82-3407; fax: �/49-

7247-82-3817

E-mail address: [email protected] (U. Fischer).

Fusion Engineering and Design 63�/64 (2002) 493�/500

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are two distinctive features of the IFMIF neutronsource which require special attention and exten-

sive development work: (1) The simulation of the

source neutron generation by Li(d, xn) reactions

in the neutron transport calculation, and, (2) the

extension of the neutron source spectrum up to :/

55 MeV. Thus neutron cross-section data for

neutron transport and activation calculations

must be provided above 20 MeV, which is theupper energy limit of the standard nuclear data

libraries.

This paper presents an overview of the R&D

work conducted in the framework of the IFMIF

project at Forschungszentrum Karlsruhe (FZK) in

co-operation with the Institute of Nuclear Power

Engineering (INPE), Obninsk, and the University

of Wisconsin (UW), Madison, on the developmentof the required computational tools and nuclear

data above 20 MeV neutron energy. The focus is

on the progress achieved recently for the D�/Li

neutron source term modelling with an advanced

Monte Carlo procedure making use of newly

evaluated double-differential data for the6,7Li(d, xn) reactions, and the creation of the

Intermediate Energy Activation File IEAF-2001.The paper includes major results of validation

calculations for the Li(d, xn) source term and

applications to the neutronic and activation ana-

lysis of the IFMIF High Flux Test Module

(HFTM).

2. IFMIF neutronics: methodology and nucleardata

2.1. General methodology

Dedicated computational tools and data are

required to enable neutronic calculations for the

IFMIF neutron source. These tools must be

capable of simulating the transport of neutrons

generated by Li(d, xn) reactions and of photonsproduced both in the Lithium target and the

material test assembly. In fusion technology ap-

plications, the Monte Carlo code MCNP [2] is the

preferred computational tool to handle neutron

and photon transport problems in a suitable way

[3]. The application to IFMIF neutronics requires

to add to MCNP the capability for simulating thegeneration of Li(d, xn) source neutrons. This can

be achieved by integrating analytical models to the

code or by making use of evaluated d�/6,7Li cross-

section data as described below. The Monte Carlo

calculation provides neutron and photon flux

spectra distributions which can be used for calcu-

lating nuclear responses of interest such as the

displacement damage, the gas production, thenuclear heating as well as other specific reaction

rates.

2.2. Nuclear cross-section data needs

Nuclear cross-section data must be provided

over the whole neutron energy range of IFMIF

which extends up to :/55 MeV. Such data must be

evaluated for a variety of nuclides important forneutron transport calculations. They must include

all data types and reactions which are required to

calculate the important nuclear responses. For

activation calculations, on the other hand, a full

set of data for all potential target nuclides must be

available. To allow the preparation of working

libraries for use with state-of-the-art transport and

activation codes, complete data sets must beprepared in accordance with standard nuclear

data format rules.

In the early stage of the IFMIF project, the

evaluation effort in the EU focused on the

preparation of general purpose neutron cross-

section data files for selected important nuclides

such as 1H, 56Fe, 23Na, 39K, 28Si, 12C, 52Cr, 51V up

to 50 MeV incident neutron energy to enableMonte Carlo transport calculations [4]. Data

evaluations for these nuclides have been performed

in a collaboration of FZK and the INPE, Obninsk.

Cross-section data for other nuclides such as 16O,14N, 27Al, 28,29,30Si, 31P, 40Ca, 50,52,53,54Cr,54,56,57,58Fe, 58,60,61,62,64Ni, 63,65Cu, 93Nb,182,183,184,186W became available later with the

APT project from the LANL 150 MeV evaluations[5]. Currently, a comprehensive evaluation effort

for 50 MeV neutron cross-section data is being

performed at JAERI [6].

In the EU, the recent evaluation effort was

devoted to the creation of a complete activation

data library for IFMIF activation calculations

U. Fischer et al. / Fusion Engineering and Design 63�/64 (2002) 493�/500494

(Section 5) and the preparation of general purposeneutron cross-section data files for the light mass

nuclides 6,7Li and 9Be. In addition, d�/6,7Li

reactions cross-section data have been evaluated

to provide the data base for the D�/Li neutron

source term in the transport calculation (Section

3).

3. D�/Li neutron source term

In the IFMIF lithium target, neutrons are

generated through the D�/Li stripping reaction

and various other nuclear Li(d, xn) reactions. The

neutron source generation must be represented

accordingly in the neutron transport calculation.

This requires developing a suitable D�/Li source

term model which ideally should be integrated to astandard Monte Carlo neutron transport code

such as MCNP.

The McDeLi code [7] has been previously

developed as an extension to MCNP with the

capability of representing the neutron source

term on the basis of a built-in semi-empirical D�/

Li reaction model. McDeLi can handle two beams

impinging onto the lithium target taking intoaccount different beam directions and a spatially

varying intensity distribution. Deuteron slowing

down in the lithium is described according to the

well established empirical model of Ziegler et al.

[8]. The Li(d, xn) reaction model considers as

neutron producing reactions deuteron stripping

and deuteron absorption followed by the forma-

tion of a compound nucleus with subsequentneutron emission. Adjustable parameters of the

Li(d, xn) reaction model were obtained through

numerical fits to experimental angle-energy dis-

tributions of neutron yields from thick lithium

targets, bombarded by 32 and 40 MeV deuterons.

Extensive testing of the McDeLi code against

available experimental data over the full deuteron

energy range from 5 to 50 MeV has recently shownthat McDeLi fails to reproduce the experimental

data below 30 MeV incident deuteron energy. The

high energy tail above 40 MeV cannot be repro-

duced either since the semi-empirical reaction

model does not take into account exothermic

reactions.

To overcome these drawbacks, a new approach

has been recently elaborated with the objective to

replace the semi-empirical Li(d, xn) reaction

model of McDeLi by a complete description of

the deuteron interactions with the lithium nuclei

through the use of evaluated d�/6,7Li cross-section

data [9]. The resulting Monte Carlo code ‘‘McDe-

Licious’’ is a further enhancement to McDeLi with

the new ability to sample the generation of D�/Li

source neutrons from tabulated d�/6,7Li cross-

section data. To provide the required data, a full

nuclear data evaluation has been performed for

the reaction system d�/6,7Li employing a newly

developed methodology based on diffraction the-

ory, a modified intra-nuclear cascade model and

standard evaluation techniques [10]. The evaluated

data include cross-sections for all reaction chan-

nels up to 50 MeV incident deuteron energy as well

as energy-angle distributions for the neutrons

emitted through the various 6,7Li(d, xn)-reactions.

A complete set of d�/6,7Li cross-section data was

prepared in standard ENDF-6 data format and

processed with the ACER module of the NJOY99

code [11] for use with McDeLicious.

The McDeLicious approach was extensively

tested against available experimental thick lithium

target data. Calculated and measured total neu-

tron yields are compared in Fig. 1 as a function of

incident deuteron energy. A comparison of angu-

lar neutron energy spectra is shown in Fig. 2 for 32

MeV incident deuteron energy. There are included

calculation results obtained with the semi-empiri-

cal D�/Li reaction model of McDeLi and the

ISABEL intra-nuclear cascade model of the high

energy particle Monte Carlo code MCNPX 2.1.5

[12]. It is revealed that McDeLicious can predict

both the neutron yield data and the angular energy

spectra with considerably better accuracy than

McDeLi and MCNPX. As McDeLicious makes

use of evaluated data files, the accuracy of the

D�/Li neutron source term can be steadily im-

proved by improving the d�/6,7Li cross-sections.

Currently new thin and thick lithium target

experiments are underway at various laboratories

[13,14] that may be used to further improve the

d�/6,7Li cross-section data.

U. Fischer et al. / Fusion Engineering and Design 63�/64 (2002) 493�/500 495

4. Neutron flux spectra and nuclear responses

McDeLicious sample calculations have been

performed for the IFMIF HFTM to demonstrate

its computational capabilities for providing the

neutron flux distribution and the nuclear re-

sponses in the material specimens. The HFTM is

a rectangular steel container of 20 cm width, 5 cm

height and 5 cm depth housing the material

specimens made of the low activation steel Euro-

Fig. 1. Measured and calculated thick lithium target neutron yields [19�/22].

Fig. 2. Measured and calculated thick lithium target neutron yield spectra at 32 MeV incident deuteron energy.


fer. The specimens will be subjected to a displace-ment damage accumulation of 20�/50 dpa (iron)

per full power year to simulate the radiation load

of the highest loaded structural material compo-

nents of a future fusion power reactor.

The geometrical model of the HFTM consists of

a rectangular Eurofer steel box with a mass density

of 6.24 g/cm3 and the proper dimensions divided

into small cubic segments of size 0.5�/0.5�/0.5cm3. Neutron flux spectra were calculated with

McDeLicious for each of the 1000 segments of one

HFTM quadrant. Two 125 mA deuteron beams

are assumed in the McDeLicious calculation im-

pinging on the lithium target at an angle of 108with respect to the horizontal center plane, see Fig.

3. Neutron flux spectra as calculated for the

HFTM with both McDeLicious and McDeLi arecompared in Fig. 4 to a typical fusion reactor

spectrum. Note the high energy tail above 40 MeV

of the IFMIF spectrum which is due to exothermic

Li(d, xn)-reactions not taken into account by the

semi-empirical reaction model of McDeLi. Fig. 5

shows a contour plot of the displacement damage

accumulation in iron as calculated for the HFTM.

Displacement damage rates up to some 50 dpa perfull power year can be achieved.

5. Activation

To perform activation calculations for the

IFMIF D�/Li neutron source requires (i) a com-

plete activation data library comprising all targetnuclides that may be present in the materials to be

irradiated and taking into account all activation

and transmutation reactions that may occur over

the whole neutron energy range from 55 MeV

down to thermal energy, and, (ii) an activation

code capable of handling the many open reaction

channels. A suitable activation data library, the

Intermediate Energy Activation File IEAF-2001[15] has been recently developed by a collaboration

of FZK and the INPE, Obninsk, as part of the

IFMIF project. The activation code ALARA (Ana-

lytical and Laplacian Adaptive Radioactivity

Analysis), previously developed at the University

of Wisconsin-Madison as an advanced computa-

tional tool for simulating induced activation in

nuclear facilities [16], has the ability to handle theIEAF-2001 activation cross section data in a

straightforward way.

The IEAF-2001 library includes 679 target

nuclides from Z�/1 (hydrogen) to 84 (polonium)

with neutron induced reaction reactions from

10�5 eV to 150 MeV incident neutron energy.

The European Activation File EAF-99 [17] served

as basis for the activation cross-section data below20 MeV neutron energy. Threshold reaction cross-

sections were evaluated on the basis of geometry

dependent hybrid exciton and evaporation models

taking into account the pre-equilibrium emission

of clusters (d, t, 3He, a) and g-rays. A new

computational approach based on diffraction

theory and a modified intra-nuclear cascade model

[10] was employed to evaluate the cross-sectionsfor the light nuclei up to Z�/12.

The IEAF-2001 data library has been prepared

in standard ENDF-6 data format making use of the

MT�/5 (neutron, anything) option with the ex-

citation functions stored in file section MF�/3 and

the product nuclide vectors in MF�/6. A 256

group GENDF-formatted (groupwise ENDF) work-

ing library has been generated with the GROUPR-module of NJOY. An IEAF-2001 CD-ROM has

been produced and is being distributed via the

NEA data bank This activation library can be

used by any activation code capable of handling

an arbitrary number of reaction channels.

Activation calculations have been performed for

the Eurofer steel specimens of the HFTM toFig. 3. Schematic HFTM model with deuteron beams and

lithium target.


demonstrate the capability and suitability of the

IEAF-2001 data library for IFMIF activation

analyses [18]. The resulting specific Eurofer activ-

ity averaged over the HFTM volume is displayed

in Fig. 6 as function of the cooling time. Only few

radio-nuclides are significantly contributing to the

activity such as Mn-54, Mn-56 and Fe-55 in the

time range up to a few years after irradiation and

H-3, C-14 afterwards. Both Mn-54, Mn-56, Fe-55

and H-3 are primarily produced through (n, p),

(n, 2n) and (n, t) activation reactions on the

natural iron isotopes whereas C-14 is an activation

Fig. 4. Comparison of a typical fusion reactor spectrum and IFMIF neutron flux spectra calculated with McDeLicious and McDeLi

for the HFTM.

Fig. 5. Contour plot of the displacement damage accumulation

(dpa per full power years) in the HFTM in x �/y , y �/z and x �/z

planes (cf. Fig. 3).Fig. 6. Specific activity of Eurofer in the HFTM as function of

the cooling time.


product of N-14. Threshold reactions in the high

energy range above 20 MeV incident neutron

energy do not significantly contribute to the

activity of the HFTM.

6. Conclusions and outlook

Substantial progress has been achieved over the

past few years in developing computational

tools and nuclear data for neutronics and

activation calculations of the IFMIF D�/Li

neutron source. With McDeLicious and ALARA

well matured computational tools are available for

Monte Carlo transport and activation calcula-

tions. The McDeLicious approach for simulating

the generation of d�/Li source neutrons may be

further developed and integrated to the high

energy Monte Carlo code MCNPX. Additional

effort is required to further improve the d�/6,7Li

data evaluations and thereby improve the accuracy

of the D�/Li source term. In particular there are

required more thin and thick lithium target

experiments for the neutron yields and spectra,

and, on the basis of these, updated reaction model

calculations for the d�/6,7Li interaction cross-

sections.

A significant number of neutron cross-section

data evaluations has already been performed and

provided in general purpose data files in standard

nuclear data format. These data evaluations need

to be validated through integral benchmark

experiments. There remains to be established a

complete general purpose data library comprising

all nuclides and reaction data types which are

required for IFMIF neutronics calculations. A

complete activation data library, the Intermediate

Energy Activation File IEAF-2001 has been

already developed and applied in IFMIF

activation calculations. The IEAF-2001 library

needs to be validated and further developed. To

this end, both differential measurements of

activation cross-sections and integral activation

experiments are required in addition to further

improve activation cross-section evaluations.

Acknowledgements

This work has been performed in the framework

of the nuclear fusion programme of Forschungs-

zentrum Karlsruhe and is supported by the

European Union within the European Fusion

Technology Programme.

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