Shutdown Dose Rate Calculations Research Articles

Shutdown dose rate calculation for fission reactors: An application of the MS-CADIS method to OPAL

There have been considerable advances of shutdown dose rate (SDDR) calculation methods for fusion related problems, however applications of these methods in the fission reactor field have hitherto been sparse. The present study attempts to bridge this gap by investigating the applicability of SDDR calculation methods for fission reactor problems. Specifically, we aim to assess and validate whether recent advances in SDDR methods can be successfully applied in fission research reactors. To this end, we estimate the shutdown dose rate distribution at the Open Pool Australian Light water reactor (OPAL) using the rigorous two step (R2S) computational method, and we compare the calculated results with the experimental data. This method utilizes a 3D reactor model implemented in the Monte Carlo N-Particle (MCNP) transport code, the AutomeD VAriaNce reduction Generator (ADVANTG) code for geometry discretization and variance reduction calculations, and the Oak Ridge Isotope GENeration (ORIGEN) inventory code for activation calculations. To ensure robustness, we employ two variance reduction techniques, Forward Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) and Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS). To the best of our knowledge, this is the first MS-CADIS method implementation for fission reactor problems. The SDDR is estimated at ten locations within the experimental hall, all situated more than 4 m away from the reactor core.The paper shows that, the experimental observations are within the lower and upper bounds of the simulation results for 4 out of 10 locations, while the remaining observations are within a factor of 7, with one significant outlier. The calculated average dose rate is within 5% of the nominal values of the experimental observations for 3 locations. The computational results are within statistical uncertainty by using two different variance reduction techniques, with significant computational advantage of MS-CADIS over FW-CADIS for SDDR calculations. The results indicate that the combination of SDDR distribution maps, estimated dose rate energy dependance, and activation information are powerful tools in identifying the radioisotopes and reactor components dominating the SDDR. These results can contribute to better radiation safety practices in contaminated areas, by enabling the minimal dose path planning or by improving radiation shielding.

Shutdown dose rate calculations in high-temperature gas-cooled reactors using the MCNP-ORIGEN activation automation tool

Background High-temperature gas-cooled reactors (HTGRs) have many distinct features from the current light water reactor (LWR) fleet, and potential decommissioning strategies should take them into consideration. The proper characterization of the shutdown dose rates can establish a suitable strategy. Methods This article introduces a new shutdown dose rate calculation capability that relies on the MCNP-ORIGEN activation automation tool and the MCNP repeated structures to explicitly model the TRistructural ISOtropic (TRISO) particles as decay radiation sources. Results Three exercises are conducted with this capability, and their results discussed. The first two exercises verify and demonstrate the workflow. The third exercise proposes a decommissioning strategy for a high-temperature gas-cooled microreactor and studies its feasibility from the shutdown dose rate standpoint. The calculations yield a dose rate map above at the reactor citadel after a three month cool-down, showing values above 40 mSv/h. Conclusions The findings imply that the decommissioning strategy based on placing the reactor in a safe shutdown state and allowing it to cool down for three months before removing various components and extracting the fuel assemblies, may not be sufficient to minimize unnecessary exposure of personnel.

Development and validation of fully open-source R2S shutdown dose rate capabilities in OpenMC *

We present the first fully open-source capabilities for shutdown dose rate (SDR) calculations of fusion energy facilities based on the Rigorous 2-Step (R2S) methodology. These capabilities have been implemented in the OpenMC Monte Carlo particle transport code, building on its existing capabilities while also leveraging new features that have been added to the code to support SDR calculations, such as decay photon source generation. Each of the individual physics components in the R2S workflow—neutron transport, activation, decay photon source generation, and photon transport—have been verified through code-to-code comparisons with MCNP6.2 and FISPACT-II 4.0. These comparisons generally demonstrate excellent agreement between codes for each of the physics components. The full cell-based R2S workflow was validated by performing a simulation of the first experimental campaign from the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from the Shielding INtegral Benchmark Archive and Database (SINBAD). For short cooling times, the dose calculated by OpenMC agrees with the experimental measurements within the stated experimental uncertainties. For longer cooling times, an overprediction of the shutdown dose was observed relative to experiment, which is consistent with previous studies in the literature. Altogether, these features constitute a combination of capabilities in a single, open-source codebase to provide the fusion community with a readily-accessible option for SDR calculations and a platform for rapidly analyzing the performance of fusion technology.

Shutdown dose rate calculations in high-temperature gas-cooled reactors using the MCNP-ORIGEN activation automation tool

Background High-temperature gas-cooled reactors (HTGRs) have many distinct features from the current light water reactor (LWR) fleet, and potential decommissioning strategies should take them into consideration. The proper characterization of the shutdown dose rates can establish a suitable strategy. Methods This article introduces a new shutdown dose rate calculation capability that relies on the MCNP-ORIGEN activation automation tool and the MCNP repeated structures to explicitly model the TRistructural ISOtropic (TRISO) particles as decay radiation sources. Results Three exercises are conducted with this capability, and their results discussed. The first two exercises verify and demonstrate the workflow. The third exercise proposes a decommissioning strategy for a high-temperature gas-cooled microreactor and studies its feasibility from the shutdown dose rate standpoint. The calculations yield a dose rate map above at the reactor citadel after a three month cool-down, showing values above 40 mSv/h. Conclusions The findings imply that the decommissioning strategy based on placing the reactor in a safe shutdown state and allowing it to cool down for three months before removing various components and extracting the fuel assemblies, may not be sufficient to minimize unnecessary exposure of personnel.

Development of Monte Carlo hybrid parallel algorithm and application in high resolution spatial-energy flux distribution calculation

Monte Carlo transport calculations for fusion reactors are very challenging due to factors such as large radiation decay gradients, complex geometric structures, and strong neutron anisotropy. For such deep penetration shielding calculation problems, in order to obtain accurate global flux distribution, variance reduction methods and parallel computing are necessary means. When using the global weight window variance reduction technique, the huge deviation of the particle arrival weight and the weight window can lead to an extreme number of splits, which is the long history problem. If a process encounters the long history problem in parallel computing and takes a lot of time to complete, other processes will stop computing because the histories allocated in the computing cycle have been completed, seriously degrading computing efficiency. In addition, the calculation of the shutdown dose rate, a key issue in fusion neutronics calculations, requires a high “space-energy” resolution, which brings huge challenges to computational memory, especially in parallel computing. In order to overcome the degradation of parallel efficiency caused by the long history problem and the memory limitation in large-scale shutdown dose rate calculation, a hybrid parallel algorithm based on shared memory was researched. Finally, the hybrid parallel architecture was built in the Monte Carlo code cosRMC to realize dynamic load balancing between threads. Through the memory sharing among all threads of a single node, the pressure of memory consumption can be effectively relieved. In order to verify the effectiveness of the hybrid parallel algorithm developed in this paper, it was applied to the calculation of CFETR. The results showed that the hybrid parallel algorithm can effectively alleviate the impact of the long history problem on the parallel efficiency. In terms of saving memory, the hybrid parallel algorithm also showed a significant effect, which provides a guarantee for large-scale shutdown dose rate calculation.

中子活化程序ABURN在核装置停堆剂量率计算中的应用研究

中子活化程序ABURN在核装置停堆剂量率计算中的应用研究

Uncertainty Quantification of the Decay Gamma Source in Mesh-Based Shutdown Dose Rate Calculations

In fusion energy systems, part of the design effort is dedicated to the assessment of the shutdown dose rate (SDR) due to the decay photons that will be emitted from activated components. Monte Carlo transport codes are often used to obtain the neutron flux distribution in the problem domain. The neutron flux distribution is used in the rigorous 2-step (R2S) workflow to obtain the photon emission density distribution of decaying radionuclides. The photon emission density is then used as an input for a dedicated photon transport step to calculate the SDR. In this paper, the uncertainty of the decay gamma source due to the uncertainty of the neutron flux distribution in the R2S workflow is investigated. A scheme is developed to estimate the uncertainty of the decay gamma source, building on the concept of groupwise transmutation and using standard error propagation techniques. The applicability of the newly developed scheme is then demonstrated on one of the conceptual designs of the fusion nuclear science facility.

Implementation and application of PyNE sub-voxel R2S for shutdown dose rate analysis

PyNE R2S is a mesh-based R2S implementation with the capability of performing shutdown dose rate (SDR) analysis directly on CAD geometry with Cartesian or tetrahedral meshes. It supports advanced variance reduction for fusion energy systems. However, the assumption of homogenized materials of PyNE R2S with a Cartesian mesh throughout a mesh voxel introduces an approximation in the case where a voxel covers multiple non-void cells. This work implements a sub-voxel method to add fidelity to PyNE R2S with a Cartesian mesh during the process of activation and photon source sampling by performing independent inventory calculations for each cell within a mesh voxel and using the results of those independent calculations to sample the photon source more precisely. PyNE sub-voxel R2S has been verified with the Frascati Neutron Generator (FNG)-ITER and ITER computational shutdown dose rate benchmark problems. The results for sub-voxel R2S show satisfactory agreement with the experimental values or reference results. PyNE sub-voxel R2S has been applied to the shutdown dose rate calculation of the Chinese Fusion Engineering Testing Reactor (CFETR). In conclusion, sub-voxel R2S is a reliable tool for SDR calculation and obtains more accurate results with the same voxel size than voxel R2S.

Development of unstructured mesh tally capability in cosRMC and application to shutdown dose rate analysis

Accurate estimation of shutdown dose rate (SDDR) is one of the key issues in the neutronics calculation of fusion reactors. cosRMC R2S was developed based on the Rigorous two-step (R2S) method using a mesh tally for shutdown dose rate calculation, which requires higher transport calculation accuracy in complex geometry. However, the traditional method in this approach was based on the orthogonal structured mesh. The orthogonal structured mesh does not necessarily coincide with geometry and material boundaries, which can introduce large errors in activation calculation and reduce the calculation accuracy of shutdown dose rate. To solve this problem, the unstructured mesh tally capability was developed in the Monte Carlo code cosRMC. On this basis, cosRMC R2S was further developed, using unstructured meshes to replace structured meshes. Due to the geometry particularity of the unstructured tetrahedral mesh, the rejection sampling strategy was adopted to sample the position of the decay gamma source. In order to verify the reliability and validity of the code, the verification analyses were carried out on the ITER port plug benchmark and the FNG ITER shutdown dose rate benchmark respectively. The results obtained with unstructured meshes were compared with those obtained with structured meshes, and the reliability of the code was preliminarily verified. The results showed that the maximum discrepancy between experimental results was 10.1% with unstructured meshes, which was closer to experimental results than that with structured meshes, indicating that the computational accuracy of the result was improved effectively. It is demonstrated that the code developed in this paper is capable of accurate shutdown dose field calculations and has the potential to be applied to complex shielding devices.

Shutdown dose rate assessment of the interspace components of wide angle viewing system diagnostic for ITER

Shutdown dose rate (SDDR) calculations have been performed to give support to the design of the Interspace (IS) VIS/IR WAVS diagnostic components of the Equatorial Port 12 (EP12), in the framework of the F4E-FPA-407 SG04 project. The Vis/IR WAVS comprises the viewing systems in four Equatorial Ports: 3, 9, 12 and 17, reminding that the present results are relative to the EP12 specifically. The Vis/IR WAVS is an optical diagnostic aimed at monitoring in visible and infrared radiation the ITER plasma facing components for machine protection.This paper describes the methodology used and shows the results obtained. The objective of the assessment is to evaluate whether the design meets the ALARA criteria in the IS area. In this work, once the materials have been selected to contribute as less as possible to the activation, the main criterion is that the contribution of all diagnostics installed in the port should not overcome the target of the biological dose rate of 100 μSv/h in the IS area after 106s of decay time.The main conclusion of the analysis is that the contribution of the VIS/IR WAVS IS components to the SDDR meets with the ALARA criterion in the IS. The contribution of the VIS/IR WAVS IS components to the SDDR in the IS, concerning the contribution of the whole ITER model, is about 9.6% in the left corridor and 5.5% in the right corridor. Besides, the percentage of contribution of the SS316L structure with respect to the total SDDR induced by the IS VIS/IR WAVS components is 99.3% in the left corridor and 99.5% in the right corridor. However, it has to be taken into account that these values have been obtained using a local model, therefore these values will be lower than the realistic values. Therefore, these values should be taken only as relative values.

Development of shutdown dose rate calculation code based on cosRMC and application to benchmark analysis

Accurate estimation of shutdown dose rate (SDR) has important reference significance for the design, operation and maintenance of fusion devices. In order to obtain the decay gamma source distribution with higher spatial resolution, a shutdown dose rate calculation code was developed based on the Rigorous two-step (R2S) method. The Monte Carlo code cosRMC and the radioactivity inventory code ALARA were coupled to calculate SDR in complex geometry based on mesh-tally. The ray tracking method was used to obtain the voxel-averaged nuclide density, and a decay photon sampling module was developed to accurately describe the distribution of the activation source. In order to improve the computational efficiency in deep penetration problems, the global variance reduction technique based on iterative weight window was adopted. The developed code was firstly verified with rectangular parallelepiped model consists of steel and water, in which the effect of bremsstrahlung photon was also discussed. Then the code was applied to the calculation and analysis of the ITER port plug benchmark, by comparing with other SDR calculation codes. The influence of different activation libraries on the decay gamma spectra and the effect of mesh resolution on the calculation accuracy were investigated. The results showed that the SDR function in cosRMC code is able to calculate SDR accurately with high spatial resolution for deep penetration problems.

Shutdown dose rate calculations with modified DEMO single sector model

A new approach for particle transport was used to calculate the Skyshine dose rates for the DEMO tokamak for the unsymmetrical case with one breeding blanket module removed for maintenance. Transport calculations for the large tokamaks, including DEMO and partly ITER, are usually performed in sector models postulating cyclic symmetry of geometry and source terms and applying reflecting boundary conditions on the sector edges. In the case of non-symmetric localized sources the sector models are not suitable.The calculation of shutdown dose rates, caused by one single tritium breeding blanket module moved from DEMO vacuum vessel into the upper maintenance hall, is an example of a localized radiation source and cannot be calculated in the standard way using a sector model. A new approach was followed for the calculations using the same sector model of DEMO, however with an according modification of the MCNP transport code. In this article the new approach is presented including its verification based on a simpler baseline model of DEMO. The results show that the obtained radiation dose distribution due to the one moved breeding blanket module differs by a factor of up to 2.8 from what would be obtained using a single sector model.

Shutdown dose rates calculations due to light ions induced activation using D1S methodology

One of the main radiation sources in nuclear facilities, which cause exposure of workers conducting maintenance activities, is the decay of radioactive nuclides. The determination of these radiation fields usually requires both radiation transport and activation calculations. The Direct-one-Step methodology allows obtaining response functions associated to the residual activity with a single Monte Carlo transport calculation. This methodology has been widely used in the analysis of nuclear fusion installations like ITER, JET or DEMO, proving to be a very efficient tool for Shutdown Dose Rate calculations. Yet, this methodology has only been applied when the material activation is due to the interaction with a neutron flux. Since, typically, in particle accelerators the D1S assumptions are met, it is proposed in this work to apply the D1S methodology to material activation induced by light ions interactions, namely: protons and deuterons.This methodology has been implemented in the D1SUNED code. In order to show its performance, the D1SUNED has been used to calculate the SDR in the LIPAc geometry, resulting in a good convergence with a reference calculation.

APPLICATION OF THE JSIR2S CODE PACKAGE FOR SHUTDOWN DOSE RATE CALCULATIONS ON JET

In this paper we present a computational exercise for shut-down dose rate calculations for the JET tokamak using the in-house developed JSIR2S code package as part of its validation. The computation is performed in two parts: neutron transport and transport of secondary gamma radiation. In order to calculate neutron activation reaction rates with sufficiently low variance, hybrid variance reduction techniques using the ADVANTG code have been utilized. Probability based sampling of secondary source particles was performed. Calculated gamma dose rates after shut down are compared with dose rate measurements performed on site using ionization chambers. The C/E agreement for 1st octant is between 0.8 to 1 while statistically meaningfull results for the 2nd octant are yet to be obtained.

Shutdown dose rate benchmarking using modern particle transport codes

Shutdown dose rate calculations provide an essential input to the design and research of fusion power plant technology. They allow the estimation of dose to personnel and equipment during planned and unplanned maintenance. The mesh coupled rigorous 2 step (MCR2S) methodology used at Culham Centre for Fusion Energy (CCFE) was originally developed to link the MCNP particle transport code and the FISPACT-II inventory code. As new particle transport codes are developed there is a strong motivation to move towards a code agnostic approach. This paper details the integration of MCR2S with the FISPACT-II API and two other transport codes, Serpent 2 and OpenMC. Two benchmarks, the FNG shutdown dose rate experimental benchmark and the ITER computational benchmark, have been performed and compared to results produced with MCNP. In general, the results show that MCNP, Serpent 2 and OpenMC give shutdown dose rate results similar to the FNG experiment for both experimental campaigns. However, all codes appeared to slightly overestimate the dose rates for Campaign 1 (all results had a C/E between 1 and 1.5) and underestimate the dose rates for Campaign 2 (all results had a C/E between 0.6 and 1). Differences were seen between OpenMC and MCNP for the ITER port plug benchmark, where the lack of variance reduction in OpenMC meant that the neutron flux estimates at the rear of the model were not converged. This led to differences of up to 13% in the shutdown dose rates. It was shown that Serpent 2 and MCNP, where variance reduction was used, gave shutdown dose rates within 3% of each other. Although some areas for development of the Serpent 2 and OpenMC transport codes have been highlighted, overall the comparisons give confidence that the implementation of these two transport codes into the MCR2S work-flow has been carried out successfully.

Engineering Calculations and Preparation for Manufacturing of ITER Equatorial Port #11

The current integration state of ITER Equatorial Port #11 (EP#11) is described. This port is the first plasma port and thus the integration phase is close to the final stage. The new modular design of neutron shielding is implemented in the Diagnostic shielding modules’ (DSM) construction to achieve the maximum shielding efficiency within the weight restrictions and keeping fully operable the tenant diagnostics systems. The same approach was applied to Interspace structure frame design to allow the safe hand-on maintenance near the highly activated components. The current iteration of the engineering calculations, including the neutron transport and shutdown dose rate (SDDR), thermohydraulics, electromagnetics, seismic, and structure integrity calculations, showed that the current design of EP#11 satisfy all ITER requirements. The special metalworking, welding, and various other technologies are tested during the process of preparation for the manufacturing of ITER EP#11. Production technologies of the full-size DSM mock-up and various elements of its infrastructure were tested. The multizoned clean area assembly hall for the final assembly is close to getting completed and the testing of vacuum items has to be sent to ITER.

Progress in development of advanced D1S dynamic code for three-dimensional shutdown dose rate calculations

Abstract The Advanced Direct 1-Step (AdD1S) is one of the most validated tools for the evaluation of Shutdown Dose Rate in complex fusion tokamak machines. The present work is built on the experience in analyzing tokamaks using prior versions of the code, where the activation analysis was limited to the first-step reactions. In the light of the neutron irradiation scenarios foreseen for fusion power reactors and the increasing importance of safety requirements, the possibility to treat the multi-step reactions introduced in the present version represents a novel and relevant development in the field. Moreover, a new Python library, PyD1S, has been developed to optimize the MCNP-FISPACT interface and data analysis. In this paper, the recent developments and the first applications are presented.

Shut-down dose rate analysis for the activated target assembly in IFMIF-DONES during maintenance

• The activation and shutdown dose analyses of the IFMIF-DONES. • Transport and activation calculations using McDeLicious code. • The dose rate maps in the access cell as well as the shielding assessments is presented. • The analysis of the shielding for activated target of DONES. This paper presents the activation and shutdown dose analyses of the IFMIF-DONES (International Fusion Materials Irradiation Facility – DEMO-Oriented NEutron Source) Target assembly. To this end, a series of coupled transport and activation calculations are performed using McDeLicious code, which is based on the MCNP Monte Carlo code for simulating the neutron production from d-Li reactions, and the FISPACT inventory code. The latest DONES test cell model has been employed and FENDL neutron cross-section data library has been used. Shut-down dose rate calculations were performed using the R2Smesh code system developed at KIT. The obtained dose rate maps in the access cell as well as the shielding assessments will be presented in the paper.

Shutdown dose calculations for the IFMIF-DONES lithium loop cell using variance reduction techniques

Abstract Shutdown dose rate (SDR) calculations for the lithium loop cell (LLC) of the IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) have been performed in this work. A new version of Rigorous 2-Step code R2Smesh-3.0 was utilized, which shows improvements on efficiency, robustness, and reliability in SDR simulations. The quality of the dose map in the LLC has been improved using a weight-window mesh, combining with an intensity-based source sampling technique. The dose rate in most of the area of the LLC is below 0.5 μSv/h at one day after shutdown. Strong gamma streaming through the empty lithium pipe is a possible concern for the lithium loop maintenance.

Benchmark analysis of coupled neutron/decay-gamma transport method NASCA for shutdown dose rate calculation

The NASCA method was developed with coupled neutron/decay-gamma transport scheme for the shutdown dose rate (SDDR) estimation in fusion device. Recently it has been upgraded for the purpose to quantify the dominant nuclides, which results in the production of radionuclides due to the neutron irradiation. In this paper, the upgrading work was put forward through planning to change the flow of the activation process of material. The linear factor between mono-energetic neutron flux and decay gamma intensity has been calculated on each source nuclide substituting compound material in previous NASCA. By using this method, the source nuclides can be singled out according to their contribution to the final dose rate results. The upgraded NASCA method has been implemented within cosRMC particle transport code and validated through ITER shutdown dose rate benchmark analysis. The results have been compared with other R2S codes. It showed the new function in NASCA is available to decide the dominated source nuclide contributors as well as to calculate shutdown dose rate accurately.