Shutdown Dose Rate Research Articles

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.

Assessment of Activation on Level L3 of the Tokamak Building due to the ITER Tokamak Cooling Water System

The ITER fusion reactor is being built to demonstrate the feasibility of fusion power and will be the largest tokamak in the world. The tokamak cooling water system (TCWS) will extract the heat generated during operations and includes large amounts of piping and equipment such as pumps and heat exchangers (HXs) that are located in a large shielded region on level L3 of the tokamak building. During operation, water in the TCWS will be activated by plasma neutrons and then flow into this shielded region. The activated coolant will in turn activate the steel in the TCWS during operation and result in an activation gamma source and radiation responses that must be assessed to inform equipment selection and maintenance schedules.The activation of materials in the shielded region of level L3 was assessed at several decay times and for different equipment options using the Oak Ridge National Laboratory (ORNL) shutdown dose rate (SDDR) code suite. The ORNL SDDR code suite implements the rigorous two-step method using the Multi-Step Consistent Adjoint-Driven Importance Sampling (MS-CADIS) method to create effective neutron variance reduction parameters for the photon response of interest. Two different HX designs, shell and tube and shell and plate, were considered, as well as the impact of cobalt impurities in steel equipment.

A Review on Noble Metals in Controlling Intergranular Stress Corrosion Cracking in BWRs

Intergranular stress corrosion cracking (IGSCC) is common in boiling water reactor (BWR) components. Corrosion problem is a serious matter that has overwhelmed the light water reactor (LWR) industry for many years. The conditions in which IGSCC takes pace due to stress, a sharpen microstructure and an environment that will facilitate the cracking while injecting H2 into the feed water system. Nuclear reactor made up of stainless steel facing serious Intergranular stress corrosion cracking due to the injection of hydrogen dosage into the nuclear reactor in the form of heavy water chemistry. Moderately large concentrations of H2 may be essential in nuclear power plants to bring the ECP below the critical value of –230 mV (SHE) to ease Intergranular stress corrosion cracking, which in turn results in an increase of steam line dose rate and shutdown dose rate and hence heavy water chemistry is limited in the nuclear reactor plant. Hence this led to the development of the concept called noble metal chemical addition (NMCA) to control IGSCC even in the presence of H2. This review gives the insights and development of NMCA on corrosion control in BWRs.

Neutronic effects of the ITER Upper Port environment update in C-model

This paper presents analysis of neutronic effects carried out with the newly developed MCNP model of the ITER Upper Port (UP) environment integrated into the neutronics C-Model of the ITER tokamak machine. C-Model includes only the standard ITER components and generic port plugs. With a number of other important updates in addition to the description of the UP environment, the new C-Model R180430 has been issued 30/04/2018, and currently is the ITER-reference model to be used in any ITER-related neutronics analysis. One of ITER’s design targets is to maintain a limit of 100 μSv/h for Shut-Down Dose Rate (SDDR) after 12 days of cooling time in the areas of envisaged personnel access with hands-on maintenance. In case of UP, the maintenance access is planned inside of UP duct where Inter-Space Structure (ISS) will be installed. Therefore, neutronics analysis has been focused on estimations of SDDR inside the UP duct and, following the ALARA principle, on finding the UP design solutions for reduction of dose. This task has been accomplished by 3D visualization of the SDDR map at UP duct, and drawing SDDR isosurfaces which revealed areas of dose hot spots and the radiation streaming pathways. Such information allows an understanding of the weak areas of the UP shielding and the formulation of the means for potential improvement of the radiation conditions in the UP duct area. The R2Smesh code has been used in SDDR calculations, with capability to plot distributions of decay gamma sources and revealing which radioactive materials are dominant contributors to SDDR. Comparison with the previous version of the ITER generic model called C-lite V2R150304MOD shows that the newly created C-Model R180430 demonstrated similar SDDR at the UP ISS maintenance corridors.

Progress in nuclear analyses of the ITER TBM Port Plug with Dummy TBMs

To achieve the validation and testing of tritium breeding blanket concepts, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are tested in three equatorial ports of the ITER tokamak. The TBM Port Plugs consist of Frame and TBM-Sets, which will be tested sequentially, each one tailored to the specific plasma operation scenario and test objective. In case a TBM-Set is not available, it can be replaced by a Dummy TBM at any time. TBM Frame and Dummy TBM are made of stainless steel and also need to meet ITER requirements on shielding, heat removal and safety, including activation. The maintenance operations for the TBM Port Plug include various activities in the Port Cell as well in the Hot Cell. In order to assess the requirements, a detailed MCNP model of the TBM Port Plug (PP), the Pipe Forest (PF) inside the Port Interspace (PI) and the Bioshield Plug (BP) was developed and integrated into the ITER neutronics “C-Model”.This analysis concerns the neutronic behaviour of the TBM PP with two Dummy TBMs by means of activation and shut-down dose rate (SDDR) calculations. The analysis includes the evaluation of the radiation fields around the TBM PP inside the ITER tokamak at several cooling times. Two configuration scenarios have been considered, one with a PF and PF structure inside the PI and one without this structure. The results show that the contribution to the SDDR from the TBM PP at 12 days after reactor shut down is small compared to the contribution originating from the PF structure.

Neutronics analysis for ITER UP18 preliminary design

The nuclear heat and the shut-down dose rate (SDDR) in the ITER upper port 18 (UP18) were estimated to provide the nuclear heat load for the structural analysis of UP18 and to provide the basis for the further SDDR mitigation strategy of UP18. In addition, the contribution of UP18 to the coil nuclear heating was also estimated to ensure the safe operation of superconducting coils. The UP18 MCNP model has been developed based on the actual CAD model, which was integrated into C-Model, the global MCNP model for ITER. While ITER UP18 accommodates three tenant systems of VUV(Vacuum Ultra-Violet)-edge Spectrometer, Neutron Activation System (NAS), and Upper Vertical Neutron Camera (UVNC), detailed models of VUV-edge and NAS were included in the UP18 MCNP model and UVNC was represented as conceptual stainless steel boxes due to lack of detailed model. For SDDR estimation, overall SDDR at the UP18 interspace was estimated as well as the SDDR contributions from tenant systems. The SDDR result includes the crosstalk from neighboring ports which are filled with general port models included in C-Model. Several SDDR mitigation options were also evaluated such as to use low Co contents material, to place B4C shielding at the rear box of port plug and to use lead shielding at the interspace floor. The results indicate that evaluated options can contribute to the mitigation of SDDR in interspace, and especially to place B4C shielding at the rear box region of port plug structure can effectively mitigate the SDDR behind the closure plate.

Standardized integration of ITER diagnostics Equatorial Port Plugs

The Diagnostic Shielding Modules (DSM) are secondary containers of the ITER diagnostic Port Plugs where shielding and diagnostics components have to be integrated together. For the development of Equatorial Port DSMs several key requirements have to be met by the Port Integrators. First of all, the total dry Plug weight must not exceed the allowable weight (currently 48 ton) in order to guarantee the consistency with the specification of other interfacing systems involved in the Port Plugs maintenance. Secondly, a shielding capability able to limit Shut-Down Dose Rate in the Interspace Zone (man access corridors) below 100 μSv/h needed to allow the human accessibility for the Port Plug sealing disassembly, the inspection and the maintenance of diagnostic components behind the Plug closure plate and in the interspace. Finally, Remote Handling (RH) compatibility for refurbishment of the Port Plug and maintenance of systems integrated is necessary as well. These operations involve activated components that are processed in the Hot Cell. As a result of these demands, the ITER Port Integration team has developed the so called Modular Equatorial DSM able to answer the weight/shielding/RH requirements described above following the ALARA principle and offering a standard infrastructure for the common DSM integration problems.

Uncertainty propagation from neutron flux to decay gamma source in R2S methodology

A new methodology to propagate the uncertainties from the neutron flux to the decay gamma source in R2S calculations is presented. The calculation of the gamma source uncertainty is an intermediate step in the propagation of the neutron flux uncertainty to the shutdown dose rate calculated using the R2S approach. The estimation of such uncertainty is not carried out in the usual implementation of current R2S tools, in spite of being an important issue of the R2S methodology.The methodology presented propagates the neutron flux uncertainty through intermediate quantities, like the reaction rates and isotope activity, to the gamma source uncertainty. In a first approach, the activation step is performed considering one step pathway.The tool based on this methodology and developed to calculate the gamma source uncertainty is also presented. Besides, it can also obtain maps with the activity of each isotope and its associated uncertainties. This tool has been tested for the simple case of the activation of a slab. The results obtained are in good agreement with the expected values.

Design of the ITER EC upper launcher nuclear shielding

ITER will be equipped with four EC (Electron Cyclotron) Upper Launchers (UL) of 8 MW microwave power each with the aim to counteract plasma instabilities during operation. These launcher antennas will be installed into four upper ports of the ITER vacuum vessel. Beside their functional purpose the port plugs which are the structural system of the launchers, have to provide as much shielding as possible in order to protect adjacent components from neutrons and photons and to minimize the shutdown dose rate in the port interspace and the port cell, being located further back in the ITER Tokamak building. Thus appropriate shielding blocks shall be installed at proper positions inside the plug. In addition several structural components will be dressed up geometrically in order to provide maximum shielding capability. This paper presents the general design of the shielding elements and their technical integration into the EC Upper Launcher.

Shutdown dose rate calculation along the ITER IVVS port

The ITER In-Vessel Viewing System (IVVS) consists of six identical units located at the B1 level of the Tokamak complex, at lower ports 3, 5, 9, 11, 15 and 17. It can be used for visual inspection of the plasma chamber in between plasma pulses. This work is devoted to the nuclear analyses performed for the latest design of the IVVS, to assess the shutdown dose rate (SDDR) in the port cell (PC) and the gallery at 1 day after shutdown in order to verify the design limit in the maintenance area. This analysis was performed through a combined use of the R2S method, for the activation of components from the vacuum vessel VV up to the bioshield, and Advanced D1S method for activation of components from the bioshield up to the gallery. Several MCNP calculations have been performed to transport neutrons and decay gammas up to the bioshield, then, by means of secondary sources, the radiation field has been extended in the PC up to the gallery. This strategy allowed to provide high resolution (voxel size 10 × 10 × 10 cm3) maps of shutdown dose rate in the PC with a statistical uncertainty within ±1% and in the gallery with a statistical uncertainty within ±20%. The consistency of the SDDR results obtained by the two codes at the interface highlighted the reliability of the coupled techniques used in the present analyses. Moreover it is assessed that the design target of 10 μSv/h in the PC is satisfied for the IVVS port.

Validation of the MS-CADIS Method for Full-Scale Shutdown Dose Rate Analysis

Fusion energy systems present increasingly significant computational challenges as they grow in size and complexity. Once constructed, ITER will be a full-size nuclear facility with highly complicated structures and support systems, with an array of scientific equipment in close proximity to the neutron-emitting deuterium-tritium plasma. Characterization of shutdown dose rate (SDDR) distributions caused by the neutron activation of these structures is important to the final design and full-power operation of the device. This work summarizes the theoretical basis and parallel implementation of the Multi-Step Consistent Adjoint-Driven Importance Sampling (MS-CADIS) method designed specifically for highly efficient execution of multistep activation problems. Fusion SDDR benchmark problems have been solved with these new tools, and the results have been compared to experimental and other computational results to establish their validation basis.

Integration of the Full Tokamak Reference Model with the Complex Model for ITER Neutronic Analysis

The ITER International Organization has developed a number of reference Monte Carlo N-Particle (MCNP) models including the tokamak machine C-model, the Tokamak Complex model, and the neutral beam injection (NBI) systems model. The Tokamak Complex model primarily describes building structures beyond the bioshield. Representation of the tokamak and its systems are not included in this model. The Oak Ridge National Laboratory Radiation Transport Group has conducted two ITER neutronic analysis model integrations: (1) integration of the tokamak C-model with the Tokamak Complex model for shutdown dose rate characterization in Port Cell 16 at level B1, and (2) integration of the NBI model with the Tokamak Complex model for estimating the spatial distribution of biological dose rate at levels L1, L2, and L3 of the Tokamak Complex. The integrated models were further extended to include models of system components that are essential to the neutronic analyses. This paper presents the approach and computer tools used to integrate existing reference models, describes the additional design details implemented in the integrated models, and provides representative neutronic calculations based on the extended models.

Neutron shielding assessment for the Remote Handling Lower Port rack of ITER

A neutron shielding assessment of an integrated ITER C-lite neutronics baseline model for the Remote Handling Lower Port rack and port interspace components has been performed. The aim of this paper is to decrease the neutron leakage as a mean to eventually reduce the activation relevant for shutdown dose rates. The performance of the neutron shielding assembled in the baseline model of the Remote Handling Lower Port rack has been assessed. As a mitigation strategy, neutronics calculations have been conducted to evaluate the radiation field during operation in the lower port with the Remote Handling Lower Port rack and the interspace components installed. A progressive improvement of the neutron shielding has been carried out with additional shielding elements including a large shielding block. As a result, the neutron streaming has been significantly reduced along the lower port with a neutron fluence rate in the closure flange in the range of 1·109 n/cm2s to 1·108 n/cm2s.

Update in the nuclear responses of the European TBMs for ITER during operation and shutdown

The depiction of the nuclear responses of the ITER European Test Blanket Modules (TBMs), Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebbles Bed (HCPB) is presented in this work. Following important components update, and important methodological advances, the nuclear heat and the tritium production have been revisited, giving new estimations 10% higher than the previous evaluation for nuclear heat in both TBMs and to 15% higher for HCPB T production. This has an impact on the thermo-mechanical design of the TBM and the tritium handling. In addition, the Shutdown Dose Rates in the respective port interspace have been characterized in local approach. It shows a performance that could imply compatibility with planned in-situ maintenance activities when analysed in global approach, an improvement with respect to previous evaluations.

Neutronic effects of diagnostic shield module length on radiation environment of ITER diagnostic generic upper port plug

Newly developed 3D CAD-based MCNP neutronics model of ITER Diagnostic Generic Upper Port Plug (DGUPP) is a baseline MCNP model of the ITER tokamak Upper Port Plug. This paper presents 3D neutron flux and Shut-Down Dose Rate (SDDR) maps for the DGUPP integrated into the MCNP C-lite V2 global model. The SDDR assessments have been performed with the R2Smesh approach and focused on the UPP Inter-Space Structure (ISS), where personnel access is planned. Following the ALARA principle, SDDR in the ISS should be minimized, for that reason the length of the Diagnostic Shielding Module (DSM) was extended by filling all the UP structure till the UP Closure Plate with the DSM shielding material. The corresponding MCNP model was called DGUPP “long-DSM”, in contrast to the original CAD-based “short-DSM”. The results of the DSM extension indicated that the strongest, up to 3 times reduction of fluxes & SDDR could be reached at the UP Closure Plate, while inside the maintenance corridors the SDDR reduction is 11%. Found possibility to reduce the SDDR emphasizes the importance of this work.

Study on collimator design for Neutron Science Facility of RAON accelerator complex

The RAON heavy ion accelerator complex is under development for the Rare Isotope Science Project (RISP) in Korea. The Neutron Science Facility (NSF) in RAON will perform Time of Flight (TOF) experiments in order to measure cross section data of fast neutrons. Fast neutrons are produced from a graphite target irradiated by a 53 MeV of deuteron beam. Consequently, a collimator that provides a highly focused neutron beam to the TOF system while reducing the background in the TOF hall should be considered. The collimator was designed by optimizing several design parameters such as the beam channel diameter, beam channel shape, and collimator type. Additionally, for the safety of workers during the operation of the NSF, an activation analysis was carried out. As a result, the final design of the collimator for the RAON NSF has a 4 m length, and channel of 0.5/2 cm in the entry/exit radius in a conical-opened shape. Materials were arranged in a ring type by setting 1 cm of the iron inside, and 20 cm of 5% borated polyethylene outside. With the final design of the collimator, a 95.7% background reduction was achieved while maintaining the original neutron beam intensity. The shut-down dose rate inside the TOF room satisfies the exposure limit

Scoping studies of shielding to reduce the shutdown dose rates in the ITER ports

The planned in situ maintenance tasks in the ITER port interspace are fundamental to ensure the operation of equipment to control, evaluate and optimize the plasma performance during the entire facility lifetime. They are subject to a limit of shutdown dose rates (SDDR) of 100 µSv h−1 after 106 s of cooling time, which is nowadays a design driver for the port plugs as well as the application of ALARA. Three conceptual shielding proposals outside the ITER ports are studied in this work to support the achievement of this objective. Considered one by one, they offer reductions ranging from 25% to 50%, which are rather significant. This paper shows that, by combining these shields, the SDDR as low as 57Δ µSv h−1 can be achieved with a local approach considering only radiation from one port (no cross-talk form neighboring ports). The locally evaluated SDDR are well below the limit which is an essential pre-requisite for achieving 100µSv h−1 in a global analysis including all contributions. Further studies will have to deal with a realistic port plug design and the cross-talks from neighbour ports.

Fusion DEMO reactor design based on nuclear analysis

The neutronics of a fusion reactor is crucial in the conceptual design of all in-vessel components and the overall reactor design. A successful DEMO reactor must demonstrate self-sufficient production of tritium to allow sufficient plant availability and electrical generation. The neutronics shows that the design constraint of the neutron wall load (NWL) is related to the tritium breeding ratio and nuclear heating, which determine the fusion power and plasma configuration and are essential for the blanket design in a fusion reactor. To align the outlet temperatures of the blanket modules, the coolant piping system must be arranged to match the distribution of the NWL. 3D neutronics analysis ensures that the blanket design allows a self-sufficient supply of tritium. The maintenance scheme for the remote handling equipment to operate Japan’s DEMO reactor should consider the shutdown dose rate and residual heat in the vacuum vessel. Base on neutronics analysis, it is important to classify radwaste disposal and to develop radwaste management scenarios. The analyses in this paper show that the radiation exposure caused by nuclides is as low as 3.4 μSv/y when the radwaste is disposed in a shallow concrete pit. This radiation exposure level is below the shallow-land burial limit in Japan (<10 μSv/y).

Shutdown dose rate neutronics experiment during high performances DD operations at JET

A novel Shutdown dose rate benchmark experiment has been performed at Joint European Torus (JET) machine during the last high performance Deuterium–Deuterium (DD) campaign in preparation of future high Deuterium-Tritium experiment (DTE2). On-line continuous gamma dose rate measurements were performed for four months at two ITER relevant ex-vessel positions close to Radial Neutron Camera and on the top of ITER-like Antenna using high-sensitive ionization chambers. Decay gamma spectra at the shutdown were also collected using High Purity Germanium spectrometer to identify dominant radionuclides contributing to the dose. Neutron fluence measurements during operations were performed as well using activation foils assembly installed close to ionization chambers. The measurements are performed to validate shutdown dose rate tools used in ITER. This paper presents the experimental assembly and measurements analyses.

Shutdown dose rate analysis with unstructured tetrahedral element based R2S method using deterministic transport solver AETIUS

We have been developing an in-house deterministic neutron and gamma transport solver, AETIUS, which uses an unstructured tetrahedral mesh for spatial discretization such as Attila. We make the most of the spatial discretization of AETIUS for a shutdown dose rate analysis.This is similar to the MCR2S unstructured mesh method with the exception that the deterministic solver AETIUS is used for the neutron and gamma transport calculations instead of MCNP(X). AETIUS provides neutron spectra for every unstructured tetrahedral element belonging to the activation zones. Activation calculations are then carried out using FISPACT-2007, and the resulting gamma spectra of each element are used as decay gamma sources for the AETIUS gamma transport calculation to obtain a shutdown dose map.We calculated the shutdown dose rate on the ITER port plug computational benchmark, and our results are the most similar to those of the FDS R2S method.