APPLIED RADIATION PHYSICS GROUP

TECHNICAL NOTE ARP-097

July 2014

Comparison of Shutdown Dose Rate Results using MCNP6 Activation Capability and MCR2S

A. Turner1, Z. Ghani1, J. Shimwell2

1: CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK 2: Department of Physics and Astronomy, University of Sheffield, Hicks Building,

Hounsfield Road, Sheffield, S3 7RH, UK

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Table of Contents

1 Introduction................................................................................................................... 4

2 Methodology ................................................................................................................. 4

2.1 Model ....................................................................................................................... 4

2.2 Tallies ...................................................................................................................... 5

2.3 Normalisation ........................................................................................................... 7

3 Results ......................................................................................................................... 9

4 Conclusions and Recommendations .......................................................................... 14

1 INTRODUCTION

The activation of materials in fusion devices is an important quantity, particularly since the decay of such materials produces a gamma radiation source that persists after plasma operation has ceased. It is important to minimise this shutdown dose rate (SDR) in areas where maintenance access will be needed.

Traditionally, the rigorous two step (R2S) method has been used to produce activation gamma sources to determine the SDR, implemented at CCFE in the form of MCR2S as well as independently at several other fusion institutions. R2S is a well-established method of coupling the particle transport capabilities of a transport code (such as MCNP) with an activation code (such as FISPACT). A neutron transport run is performed, to obtain neutron flux and spectra in a mesh tally superimposed over the geometry. An activation calculation is then performed on each voxel of the mesh tally, in turn producing an activation gamma source to be used in a second (photon) MCNP calculation to determine the SDR.

Advantages of R2S approach:

Relatively simple to implement.

Production of gamma source permits flexibility in gamma calculation e.g. modification of geometry, or source placed in different model.

Disadvantages of R2S approach:

Mesh resolution affects SDR results. Neutron flux and material compositions are averaged over mesh voxels. High mesh resolutions required for accuracy and corresponding high memory consumption.

Due to use of a mesh source, decay photons can start in materials/regions where they were not originally produced.

Multiple calculation steps can lead to book-keeping/quality control issues.

No uncertainty propagation all uncertainty in the neutron flux information is lost (only average gamma source is used rather than a sampled one).

These are the issues considered native to the R2S methodology, and CCFE and other fusion institutions are working to develop solutions to these issues.

In the latest version of MCNP (MCNP6), the developers have added the capability to perform activation on-the-fly within the neutron transport run using the ACT card, hereby referred to the MCNP6-ACT capability. This feature utilises the CINDER libraries and was previously available in MCNPX.

Clearly this will have the advantage of a more realistic treatment, with photons being started where they were created rather than a mesh-averaged photon source. In addition, the photon SDR results would then have a statistical uncertainty that accounted for the neutron transport as well as the photon transport.

This report makes a comparison between previously obtained results using MCR2S, and results obtained using the new MCNP6-ACT capability, performed using the DEMO shutdown dose rate computational benchmark model. This model was previously used as part of an exercise to compare the various implementations of the R2S method.

2 METHODOLOGY

2.1 Model

The model used in this study is the 2008 DEMO-HCLL model by KIT, as used in the DEMO computational benchmark studies [1]. It consist of a 11.25o poloidal sector of a tokamak containing a detailed representation of the HCLL blanket modules and approximate

representations of the divertor, in-vessel shield, vacuum vessel (including port extensions), toroidal field coils and central solenoid. Reflecting boundary conditions are used to represent symmetry. Poloidal and toroidal sections are illustrated in Figure 1.

(a) (b)

Figure 1: MCNP model of 2008 DEMO-HCLL reactor showing mesh division for R2S calculations; (a) poloidal section at Y=2, (b) toroidal section at Z=2.

The above figure shows the activation mesh tallies specified for the R2S codes, not applicable for MCNP6_ACT.

Materials were assigned as per the original task specification document. The specification requires enriched 6Li in the lithium-lead material, which at the time of the original study could not be supported in MCR2S and was performed at natural elemental abundances in those calculations. The calculation was repeated here using the new version of MCR2S (version 2), which integrates with FISPACT-II and has the capability to use isotopic material descriptions.

2.2 Sources

In the case of the MCR2S based DEMO neutron calculation, the standard parametric plasma source was used. Fusion power is taken as 2.4 GW, corresponding to a DT neutron emission of 8.52x1020 n/s. In addition, there is no need to account for the sector angular extent when defining the normalisation as the parametric source accounts for the sector angle via a reduced source particle weight - thus this full sector normalisation should be used.

The irradiation time is accounted for in MCR2S, but needs to be included as a time-dependant source in MCNP6_ACT. The irradiation time to be assumed was 8 years at 50% power, followed by 60 days at 100% power. For the divertor, a different irradiation scenario was to be assumed, of 4 years at 50% power followed by 60 days at full power. This added complexity was ignored for the MCNP6-ACT run, since it was not clear how to simulate such a system, and in any case, the effect on the result was not expected to be significant for short cooling times.

For MCNP6 and the time dependent source, the parametric plasma source was used but has been modified to allow the user to input a time-dependent source (a series of source

times and probabilities) via the SI and SP cards , along with source biasing parameters (SB cards). The 8 year neutron source time was split into time bins and biased towards later times, which would be more important for gamma production at the shorter decay times of interest. MCNP6 supports both line and multigroup secondary gamma production, and due to the difficulty in obtaining reliable tally statistics (see section 2.3), a multigroup treatment was used, having been found to run over 100 times faster than line data. Since it was not feasible to obtain results using line data, the effect of utilising multigroup gamma data on the results has not been assessed in this study. In addition, the delayed neutron and delayed photon biasing functions were used to increase the sampling of these events, and the SPABI card used in an attempt to increase the population of high energy photons - though the effectiveness of these approaches was not clear.

Figure 2: MCNP time-dependent parametric plasma source and ACT cards

2.3 Tallies

Results were obtained for absorbed and biological dose rates at four different in-vessel locations and at decay times of 1 hour and 10 days. 10 cm diameter spherical cells were used, centred at the following in-vessel locations D1, D2, D3 and D4: D1 centre of vessel: (X,Y,Z) = (750, 60, 0) cm; D2 outboard equatorial midplane: (X,Y,Z) = (990, 60, 0) cm; D3 inboard equatorial midplane: (X,Y,Z) = (500, 60, 0) cm; D4 divertor region: (X,Y,Z) = (570, 60, -480) cm. The original specification called for absorbed dose rates to be obtained for steel-filled tally spheres (F6 tallies, i.e. with self-shielding), and photon flux spectra and biological dose rates for voided spheres (no self-shielding). As such, this requires two calculations, and it was decided to only run one calculation using MCNP6_ACT due to high computational demands. Results and comparisons for photon flux and biological dose rate are therefore provided in the results section, absorbed dose is not calculated using MCNP6_ACT. The gamma-to-dose coefficients to be used in the biological dose calculations are the ITER-recommended ICRP-74 in both MCNP6_ACT and MCR2S. Tallies for the MCNP6_ACT calculation required the inclusion of tally time bins. MCNP generates decay gamma photons for decay times between 0 and 1010 seconds by default, and tally time bins are required to obtain result in the time intervals of interest. Due to this methodology, it is not possible to produce a result for a specific decay time, but only a time

bin of finite width (t). Initially, calculations were performed assuming a 1% t window about the decay times of interest, for example for 1 hour, the time bin would be 3582 to 3618 seconds after shutdown. However, it was found that it was extremely challenging to obtain

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