Total Nuclear Heating Research Articles

Commissioning of the ESS large-scale 20 K helium refrigeration system for the cryogenic moderator system

At the ESS, a large-scale 20 K helium refrigeration system with a cooling capacity of 30.3 kW at 15 K, which is called Target Moderator CryoPlant (TMCP), has been designed to cool the cryogenic moderator system (CMS) where subcooled liquid hydrogen will be circulated at 1.0 kg/s through four hydrogen moderators where the total nuclear heating is estimated to be 17.2 kW for the proton beam power of 5 MW. The first phase of the TMCP installation and commissioning for the TMCP compressor skids and cold box was completed in 2019. The phase 2 installation of 300 m-long cryogenic helium transfer lines (CTLs) and a valve box located at one end of it in the Target building has been completed in the summer of 2022. Subsequently, the commissioning has been conducted until December 2022. This paper will describe the performance evaluation test results of the CTL heat load and pressure drop, the TMCP cooling capacity, and a test result of a failure action caused by instrument air failure.

Radiation Damage Analysis of FNSF Components Using McCad and MCNP

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa). Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches. This study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 ± 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH2, Mg(BH4)2, and ZrH2 could reduce the total heating on the magnet and also reduce the TBR.

Flux and Nuclear Heating Rate Calculations for CALORRE Irradiation in MITR

This work summarizes neutronic calculations performed for the CALORRE differential calorimeter specifically designed by Aix Marseille University to inform the irradiation campaign planning for testing in MITR within the framework of the CALOR-I research program. MCNP software was used to predict the neutron and gamma flux spectrum, and total nuclear heating rate to the components of the calorimeter in a variety of positions. Results were compared to evaluate spatial bias, core loading effects, optimization for axial position, and provide input data for thermal multi-physics modelling.

Neutronics and shielding design of CFETR HCCB blanket

Tritium breeding blanket (TBB) is one of the most important components of China Fusion Engineering Test Reactor (CFETR) with functions of tritium breeding, heat generating and nuclear shielding. This paper presents the progress on the neutronics and shielding design of the helium cooled ceramic breeder tritium breeding blanket (HCCB TBB) developed by the Southwestern Institute of Physics (SWIP). Based on the latest profile and layout configuration of CFETR TBB, the preliminary neutronics design, optimization and analysis of HCCB TBB have been completed. After several rounds of neutronics design and optimization, the TBR of HCCB TBB is 1.177, and the value drops to 1.101 considering the influence of auxiliary systems, which can meet the design objectives of CFETR. The total nuclear heating of HCCB TBB is 1402 MW with 1500 MW fusion power. Among all the blanket modules, blanket module #3 is subjected to the strongest neutron irradiation. Meanwhile, the overall shielding analysis for TF coils of CFETR has been performed based on the current design of HCCB TBB, and the preliminary shielding design and optimization for the weak shielding areas such as HCCB TBB, divertor and neutral beam injector (NBI) port have been carried out. After shielding optimization, the fast neutron fluence and nuclear heating of all TF coils are below the design limit. By increasing the number of source neutron histories and using some variance reduction techniques in the MCNP calculation, the relative errors of neutronics and shielding analysis results for HCCB TBB are all less than 5%.

Application of MCNP unstructured mesh in the design process of the ITER EC-UL M3 mirror

Finite element analysis (FEA) is widely used in engineering to accurately model physical systems. FEA enables multi-physics investigations to be undertaken efficiently and increases the cohesiveness of interdisciplinary engineering assessments. Thanks to the recent implementation of the FEA Unstructured Mesh capability (UM) into MCNP6, radiation transport is now able to contribute more effectively to such multi-physics investigations. Furthermore, the use of UM in MCNP allows complicated components to be modelled more accurately since it does not have the geometrical limitations of the alternate Constructive Solid Geometry method (CSG), leading to more accurate simulations and inherent results. The ITER Electron Cyclotron-Upper Launcher (EC-UL) M3 Mirror is a complex component currently under design that had required various engineering assessments that utilise FEA such as mechanical and thermohydraulic. Therefore, this neutronics assessment took advantage of the UM capability of MCNP6.2 to be consistent with the other assessments. The peak volumetric nuclear heating for this component was found to be 3.76 ± 22.0 % W/cm3 by this method. The total integral nuclear heating was found to be 6595.1 ± 0.6 W, an increase of 9.3 % on the previous design but with a slightly different deposition distribution. Crucially, the total integral nuclear heating of the CuCrZr reflector was determined to be 142.92 ± 0.07 W, a decrease of 69 % from the previous mirror design allowing the fulfilment of the mechanical code and standards.

Shielding proposal to mitigate the PFC #2 heating due to 16N decay photon radiation from the activated cooling water

In this work, the total heating of the ITER superconducting Poloidal Field Coil #2 (PFC #2) has been calculated due to 16N decay photon radiation emitted from the activated water in the Blanket Manifolds, the Chimney pipes and the Tokamak Cooling Water System (TCWS) manifolds. In order to reduce the total nuclear heating of the PFC #2, shielding measures have been proposed; a set of conceptual stainless-steel shields has been designed and inserted into the computational model of the ITER machine covering the Upper Port Stub (UPS) Chimney and the UPS Chimneybox. The results have shown a 42% reduction of the PFC #2 heating, i.e. from 275 W down to 160 W, when 10 cm thick plates have been placed on the mentioned UP structures. At the same time, the maximum specific heating of the PFC #2 has been decreased by 63%, from 40 W/m3 down to 15 W/m3. A recommendation has been given for further reduction of the PFC #2 heating by improving the on-going design of the TCWS duct.

Development of the Advanced Neutronic Analysis Model for the K-DEMO With MCNP Code

Neutronic assessment is the important foundation of Korean fusion demonstration reactor (K-DEMO) study. Previously, preliminary neutronic calculations using a simplified model have been performed [1] , and optimized structure of blanket module and shield material has been derived. Based on the previous calculation results, the advanced K-DEMO neutronic analysis model has been developed to perform more accurate and reliable neutronic assessments with the detailed configuration of vacuum vessel (VV), toroidal field (TF) coil, blanket, and shield. The advanced K-DEMO neutronic analysis model was imported into the integrated interface program by the help of Monte Carlo automatic modeling system to produce input files for the Monte Carlo neutron photon transport code analyses. The neutronic calculation on neutron wall loading (NWL) distribution, tritium breeding ratio (TBR), and nuclear heating in tokamak components was conducted for the newly developed advanced K-DEMO neutronic analysis model. In the NWL calculation, the peak was indicated at the inboard (IB) and the outboard (OB) blanket first wall with the value of 3.81 and 3.93 MW/ $\text{m}^{2}$ . The different types of models by changing gaps between the blanket modules and the thickness of separated borated steel shield blocks are calculated to check global TBR and the shield performance. The global TBR of 1.017 was achieved with the 49.7- and 76.2-cm-thick IB and OB blanket modules. The model with 1-cm-thick assembly gap between blanket modules, 22-cm-thick IB shield, 50-cm-thick OB shield, and 34-cm-thick divertor shield are considered as a reasonable shield model with the total nuclear heating in 16 TF coils and VV of 9 and 32 kW.

Performance evaluation of a developed orifice type heater for thermal compensation control at J-PARC cryogenic hydrogen system

Supercritical hydrogen with a temperature of less than 20 K and a pressure of 1.5 MPa is used as moderator material at J-PARC. Total nuclear heating of 3.75 kW is generated by three moderators for a 1-MW proton beam operation. We have developed an orifice-type high-power heater for thermal compensation to mitigate hydrogen pressure fluctuation caused by the abrupt huge heat load and to reduce the fluctuation in the temperature of the supply hydrogen to less than 0.25 K. Through a performance test, we confirmed that the developed orifice-type heater could be heated uniformly and showed fast response, as expected. Furthermore, a simulation model that can describe heater behaviors has been established on the basis of the experimental data. The heater control approach was studied using the aforementioned heater simulation model and a dynamic simulation code developed by the authors.

The Impact of Updated Cross Section Libraries on ITER Neutronics Calculations

In ITER, determination of radiation loads such as nuclear heating due to neutrons and photons (gammas) is an important part of the design process. Monte Carlo transport codes need accurate neutron and photon cross section libraries to produce accurate results. Because photon heating dominates the contribution to total nuclear heating for common materials like stainless steel and copper in several key components of ITER, the photon cross section library is particularly important. In this work, two ITER realistic benchmark calculation models are used to determine the impact on nuclear heating by the cross section library used in the calculation. The results show that the nuclear heating can be as much as 5% lower to as much as 6% higher than the nuclear heating calculated using the standard fusion neutron and photon cross section library.

Parametric study of the energy deposition inside the calorimeter measuring the nuclear heating in Material Testing Reactors

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material and two calorimetric cells. Then these measurements are used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices.This paper will present simulations with MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) to evaluate the nuclear heating inside the calorimeter during irradiation campaigns of the CARMEN-1P mock-up inside OSIRIS reactor periphery (MTR based on Saclay, France). The whole complete geometry of the sensor has been considered. The calculation method corresponds to a calculation in two steps. Consequently, we used as an input source in the model, the neutron and photon spectra calculated in various experimental locations tested during the irradiation campaign (H9, H10, H11, D9). After a description of the differential calorimeter sensor, the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements is introduced by two quantities: KERMA and energy deposition rate per mass unit. The Charged Particle Equilibrium (CPE) inside the calorimeter elements is studied. The contribution of prompt gamma and neutron is determined. A comparison between this total nuclear heating calculation and the experimental results in a graphite sample will be made. Then parametric studies performed on the influence of the various calorimeter components on the nuclear heating are presented and discussed. The studies of the influence of the nature of materials, the sensor jacket, the source type and the comparison of the results obtained for the two calorimetric cells leads to some proposals for the sensor improvement.

ITER Neutronics Modeling Using Hybrid Monte Carlo/Deterministic and CAD-Based Monte Carlo Methods

The immense size and complex geometry of the ITER experimental fusion reactor require the development of special techniques that can accurately and efficiently perform neutronics simulations with minimal human effort. This paper shows the effect of the hybrid Monte Carlo (MC)/deterministic techniques—Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS)—in enhancing the efficiency of the neutronics modeling of ITER and demonstrates the applicability of coupling these methods with computer-aided-design-based MC. Three quantities were calculated in this analysis: the total nuclear heating in the inboard leg of the toroidal field coils (TFCs), the prompt dose outside the biological shield, and the total neutron and gamma fluxes over a mesh tally covering the entire reactor. The use of FW-CADIS in estimating the nuclear heating in the inboard TFCs resulted in a factor of ˜275 increase in the MC figure of merit (FOM) compared with analog MC and a factor of ˜9 compared with the traditional methods of variance reduction. By providing a factor of ˜21 000 increase in the MC FOM, the radiation dose calculation showed how the CADIS method can be effectively used in the simulation of problems that are practically impossible using analog MC. The total flux calculation demonstrated the ability of FW-CADIS to simultaneously enhance the MC statistical precision throughout the entire ITER geometry. Collectively, these calculations demonstrate the ability of the hybrid techniques to accurately model very challenging shielding problems in reasonable execution times.

Radiation Streaming in Gaps between ITER First Wall/Shield Modules

Gaps exist between the first wall/shield (FWS) modules of ITER to allow for fitting by remote handling equipment. Simplified three-dimensional models were used at the top and mid-plane locations to analyze gap streaming. Heating, helium production, and fast neutron fluence were examined at the front of the vacuum vessel and the magnet for both straight and stepped gaps. In addition, total nuclear heating values in the inboard magnet and central solenoid were examined for straight and stepped vertical gaps and for combined horizontal and vertical straight gaps. The results show significant radiation streaming effects that are more pronounced for fast neutron fluence and helium production. Furthermore, it was found that stepping the gap significantly reduces the local peaking, but has little effect on the relative average values of radiation effects. The results also show increases up to 75% in total magnet heating at the inboard mid-plane location for a straight 2 cm combined vertical and horizontal gap.

Effect of hydrogen loop fluctuation on cold neutron performance with hybrid control system in JSNS

Supercritical hydrogen is selected as a moderator material in Japan Spallation Neutron Source Facility (JSNS). Total nuclear heating in the moderator is estimated to be 3.75 kW at 1 MW proton beam power, which provides a hydrogen cryogenic system as a heat load. The heat load changes significantly due to pulse proton beam operation. A control system to mitigate such heat load changes is required for the hydrogen cryogenic system in order to maintain stable operation and steady cold neutron generation. A hybrid control system, which consists of a heater and an accumulator, is developed and installed in the system. In order to build the optimal control method, the behaviors of hydrogen loop are analyzed with an original computer code. The analysis suggests that a stability of hydrogen system and a cold neutron beam performance by the hybrid fluctuation control system are compatible.

Three-dimensional Monte Carlo neutronic calculations of China ITER HCSB-TBM

Three-dimensional neutronic calculations were conducted to analyze the nuclear performance of the recent China Helium Cooled Solid Breeder (HCSB) test blanket module (TBM) in ITER. The Monte Carlo transport code MCNP was used with the nuclear cross-section data library FENDL/2 offered by IAEA. The analyses were based on a detailed three-dimensional HCSB-TBM geometrical model which was integrated in the ITER 20° global sector model. Key nuclear responses of the HCSB-TBM, such as the tritium generation ratio and the nuclear heating distribution, were calculated and discussed. The total nuclear heating is about 0.505 MW, and the peak heating density which appears in Li 4SiO 4 is less than 5.5 W/cm 3. The tritium production rate is about 13.39 mg/day while considering ITER duty factor of 0.22. The results demonstrate that the current design can meet the engineering requirement from the neutronic point of view.

Radiation shielding calculations for the vista spacecraft

Abstract The VISTA spacecraft design concept has been proposed for manned or heavy cargo deep space missions beyond earth orbit with inertial fusion energy propulsion. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris and with the help of a magnetic nozzle. The calculations for the radiation shielding have been revised under the fact that the highest jet efficiency of the vehicle could be attained only if the propelling plasma would have a narrow temperature distribution. The shield mass could be reduced from 600 tons in the original design to 62 tons. Natural and enriched lithium were the principle shielding materials. The allowable nuclear heating in the superconducting magnet coils (up to 5 mW/cm3) is taken as the crucial criterion for dimensioning the radiation shielding structure of the spacecraft. The space craft mass is 6000 tons. Total peak nuclear power density in the coils is calculated as ∼5.0 mW/cm3 for a fusion power output of 17 500 MW. The peak neutron heating density is ∼2.0 mW/cm3, and the peak γ-ray heating density is ∼3.0 mW/cm3 (on different points) using natural lithium in the shielding. However, the volume averaged heat generation in the coils is much lower, namely 0.21, 0.71 and 0.92 mW/cm3 for the neutron, γ-ray and total nuclear heating, respectively. The coil heating will be slightly lower if highly enriched 6Li (90%) is used instead of natural lithium. Peak values are then calculated as 2.05, 2.15 and 4.2 mW/cm3 for the neutron, γ-ray and total nuclear heating, respectively. The corresponding volume averaged heat generation in the coils became 0.19, 0.58 and 0.77 mW/cm3.

Nuclear features of the fusion ignition research experiment (FIRE)

The main nuclear features of the baseline design of fusion ignition research experiment (FIRE) have been evaluated. Critical issues were addressed and R&D needs were identified. Modest values of nuclear heating occur in the FIRE components. The total nuclear heating in the 16 TF coils during DT shots is 19 MW. The cumulative damage in the copper alloys used is very low (<0.05 dpa). However, issues of low temperature embrittlement and thermal creep at high temperatures need to be resolved by an R&D program. The radiation induced resistivity increase in the conductors of the TF coils is primarily due to displacement damage and is <20% of the unirradiated resistivity. The magnet insulator development R&D program should involve irradiation to dose levels up to 1.5×10 10 Rad with the proper mix between neutrons and gamma photons (50% gamma dose) relevant to FIRE conditions. Low levels of activity and decay heat are obtained. Hands-on ex-vessel maintenance is feasible. All components qualify as Class C low-level waste. Activation of nitrogen gas inside the cryostat produces a very small amount of 13N and 14C.

Optimization of the radiation shielding mass for the magnet coils of the VISTA spacecraft

An attempt has been made for the optimisation of the radiation shielding of a spacecraft design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris, and with the help of a magnetic nozzle. The allowable nuclear heating in the super conducting magnet coils (up to 5 mW/cm 3) is the crucial criterion for the dimensioning of the radiation shielding structure of the spacecraft. The optimized design reduced the shield mass from 600 tons to 93 and 88 tons with natural and enriched lithium, respectively. The space craft mass was 6000 tons. Total peak nuclear power density in the coils is calculated to be 5.0 mW/cm 3 for a fusion power of 17,500 MW. Peak neutron heating density is 2.6 mW/cm 3 and peak γ-ray heating density is 2.9 mW/cm 3 (all on different points). However, volume averaged heat generation in the coils is much lower, namely 0.30, 0.73 and 1.03 mW/cm 3 for neutron, γ-ray and total nuclear heating, respectively.

Radiation shielding mass saving for the magnet coils of the VISTA spacecraft

Radiation shielding structure of a design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit has been investigated. Fusion power deposited in the inertial confined fuel pellet debris delivers the rocket propulsion with the help of a magnetic nozzle. The nuclear heating in the super conducting magnet coils determines the radiation shielding mass of the spacecraft. It was possible to achieve considerable mass saving with respect to a recent design work, coupled with higher design limits for coil heating (up to 5 mW/cm 3). The neutron and γ-ray penetration into the coils is calculated using the S N methods with a high angular resolution in (r–z) geometry in S 16P 3 approximation by dividing the solid space angle in 160 sectors. Total peak nuclear heat generation density in the coils is calculated as 3.143 mW/cm 3 by a fusion power of 17 500 MW. Peak neutron heating density is 1.469 mW/cm 3 and peak γ-ray heating density is 1.674 mW/cm 3. However, volume averaged heat generation in the coils is much lower, namely 74, 163 and 337 μW/cm 3 for neutron, γ-ray and total nuclear heating, respectively. The net mass of the radiation shielding for the magnet coils is 200 tonne by a total mass of 6000 tonne of the space craft.

Three-Dimensional Neutronics Analysis for the ITER Divertor Cassette Design Options

Three-dimensional neutronics calculations have been performed for the ITER divertor cassette design options to determine the nuclear parameters in the cassettes and assess the impact of streaming on vacuum vessel and toroidal field (TF) coil damage. The local nuclear parameters in the components of the reference cassette design are similar or lower than those in the cassette design option with wings. The total nuclear heating in the 60 divertor cassettes is 102 MW for both designs. Helium production levels in the vacuum vessel in the divertor region allow for rewelding. The TF coils are well protected from radiation streaming into the divertor ports.

Shielding Analyses of the ITER NBI Ports

Shielding analyses of the ITER neutral beam injector (NBI) ports have been performed using three-dimensional Monte Carlo and two-dimensional discrete ordinates Sn methods. The biological dose rates inside the cryostat after reactor shutdown are expected to be lower than design target of 100 μSv/h for the current NBI reference design with ∼60 cm thick NBI port walls. It was also observed that the total nuclear heating in the toroidal field (TF) coils satisfies the design limit of 17 kW when the port wall is 40 cm thick. The Sn calculations, performed using a rectangular model of the NBI, overestimate the dose rates at the cryostat and nuclear heating in TF coils by factors of ten and two, respectively, compared to Monte Carlo results obtained using a more accurate representation of the NBI system.