Monte Carlo Transport Code Research Articles

Conceptual design of sandwich walls for shielding against secondary neutrons using MC simulations with FLUKA

FLUKA Monte-Carlo transport code was employed to evaluate the secondary neutron spectra emerging from spherical sandwich shielding configurations composed of concrete and soil, similar to that used at the particle therapy facility MedAustron. This study provides a comparative analysis of neutron spectra attenuated by a concrete-soil-concrete (CSC) sandwich wall shielding configuration versus a full concrete wall design (CCC). Furthermore, we enhanced the shielding performance of the CSC configuration by adding an additional concrete layer (CCSC) to achieve results comparable to the CCC shielding. Two scenarios were tested for shielding performance: (1) primary protons at 100 MeV, and (2) primary carbon ions (C-ions) at 190 MeV/u. Our simulations with primary protons of 100 MeV showed that adding additional internal concrete wall to the CSC configuration, therefore designing the CCSC configuration, the RP performance becomes slightly improved – the HE-peak drops from (1.43 ± 0.11)10−11 to (5.62 ± 0.3)10−12, about 2.5 times. Still, the HE-peak of the exiting neutron spectrum from CCC -(6.29 ± 1.87) 10−13 is about 9 times lower than that exiting CCSC - (5.62 ± 0.3) 10−12.Our simulations with primary C-ions showed that by placing an additional internal concrete wall to the CSC configuration (CCSC) the RP performance becomes slightly improved – the exiting HE peak can be further attenuated from (6.92 ± 0.40)10−9 for CSC to (3.79 ± 0.15)10−9, becoming comparable to the one exiting the CCC configuration, (0.92 ± 0.04)10−9, only 4 times higher. Future research should be focused on improvements of the RP performance of the CCSC, by increasing the soil layer thickness and taking into consideration the humidity (water content) in the soil and concrete and also improve the number of primaries to 109 or even 1010 for better statistical outcome.

Monte Carlo Workflow Unification for Nuclear Reactor Analysis with Metamodel-Driven Modeling

Monte Carlo (MC) transport codes are a cornerstone of nuclear reactor analysis frameworks, providing reference solutions and multigroup cross sections, and even as core components in multiphysics couplings. These applications can be seen in toolkits from the Nuclear Energy Advanced Modeling and Simulation program and the U.S. Nuclear Regulatory Commission’s BlueCRAB (Comprehensive Reactor Analysis Bundle). Contrary to their ubiquitousness in reactor physics modeling and simulation, popular MC codes, such as MCNP, Serpent, and KENO, are still reliant on antiquated textual input formats. These input languages use a plethora of keywords with terse syntax to specify all facets of a model, including its geometry, materials, physics settings, and tally options. This poses a steep learning curve and a poor user experience. Being tied to unique text-based input formats also significantly complicates programmatic input generation or modification that may be desired and/or required within a multiphysics framework. This work demonstrates how the development of programmatic interfaces for MC codes can support model unification and translation activities. Building on the Python application program interface (API) development of MCNPy, similar capabilities are being implemented for Serpent that are able to support model translation. In future work, the Serpent API capabilities will be made more robust and the work will be further expanded to include translations with KENO.

Design of a compact and portable space neutron spectrometer based on Monte Carlo

With the rapid development of space exploration, the detection of space neutron radiation is becoming increasingly important. The currently widely used Bonner sphere spectrometer have drawbacks such as large size and weight, as well as low fault tolerance, when detecting space neutron spectra. This paper describes in detail a new type of space neutron spectrometer (SNS), which has two different specifications to adapt to the directional and non-directional neutron field environment, and can measure the directional neutron energy spectrum. For the directed neutron field, SNS integrates 12 3He thermal neutron counters (diameter 3 cm: 3, diameter 4 cm: 6, diameter 5 cm: 3) and uses cylindrical polyethylene as a moderator. For non-directed neutron fields, SNS integrates 9 3He thermal neutron counters (diameter 3 cm: 4, diameter 4 cm: 3, diameter 5 cm: 2) located in a single structure made of polyethylene, boron-containing polyethylene and gadolinium. The device is capable of providing a strong directional response in the energy range of thermal neutrons up to 20 MeV, with little sensitivity to neutrons coming from directions other than the axis of the cylinder. The Monte Carlo transport code FLUKA was used to determine the final configuration of the instrument, including the arrangement, number, and position of thermal neutron counters. In addition, the response matrix of the instrument was calculated using FLUKA code. This device can replace traditional Bonner sphere spectrometer for measuring space neutrons, and it also provides reference value for downsized and lightweight neutron spectrometers on the ground.

Comparison of Monte Carlo Tools for Displacement Calculations for Protons and Heavy Ions Irradiation on Iron and EUROFER97

Abstract Ion irradiation of structural materials for the current and future generation nuclear reactors is a widespread technique used to simulate neutron irradiation in the context of material science research. During the designing phase of irradiation experiments Monte Carlo transport codes are often employed to quantify the radiation exposure, by estimating the displacements per atom (dpa) quantity. From the available displacement calculation codes, four popular choices were selected: cern-fluka, mcnp, phits, and srim. These codes will be used to simulate the irradiation effects of proton and Fe ions on pure iron (G379) and EUROFER97 alloy (a common material in fusion applications) targets. Results coming from the different software are presented in the form of damage profiles and of number of displacements as a function of the primary beam energy. This study identifies discrepancies between the codes' outputs and provides optimized simulation setup parameters.

Estimation of H*(10) and E(Ω) (and H*) using the Hp(10) and Hp(0.07) detectors of a body TLD

The H*(10) and E(Ω) responses to photons from 15 to 1253 keV were determined as a function of source orientation for the 7LiF:Mg,Cu,P ‘body’ and ‘skin’ elements of a thermoluminescence dosemeter optimized to measure Hp(10) and Hp(0.07), and are presented here relative to the results from 662 keV calibration exposures. The calculations were performed using the Monte Carlo transport code MCNP6.2, with the dosemeter located either free-in-vacuo or on a concrete wall; the impacts of the radiation field and dosemeter mounting are discussed. The relative response profiles were found to be fairly flat at energies above a few 10s of keV, leading to the conclusion that the personal dosemeter may be used in some environmental or workplace fields to provide reasonable estimates of ambient dose equivalent and effective dose; the latter is additionally relevant given the newly proposed operational quantity ambient dose (H*). Optimal correction factors were also derived that allow the dosemeter's standard Hp(d,0°,137Cs) system calibrations to be maintained.

Matrix effect correction in active neutron interrogation for residual fissile mass assessment in radioactive waste drums

Metallic waste in the form of shells and nozzles remains following the reprocessing of spent fuel at La Hague plant (France). Before the final disposal, drums containing the waste are transported to the compaction facility where its volume is reduced by a factor of 5. With the objective of controlling the criticality/safety levels, an active neutron interrogation system at the entrance of the facility is used to assess the residual fissile mass remaining in the waste, which relies on prompt fission neutron detection (known as the differential die-away technique, DDA). The measured signal is proportionally linked to the fissile mass by a calibration coefficient. However, two effects can produce an inaccurate prediction. The number of induced fissions for the same fissile mass varies according to the waste matrix composition, particularly the neutron slow-down and absorption ratio. Secondly, the presence of fissile clusters can impact the neutron interrogation due to an increase in self-shielding effect. This work presents a matrix effect correction method based on the use of internal flux monitors (sensitive to the neutron absorption ratio) and the transmission signal (sensitive to the slow-down ratio). It relies on a numerical model of the whole measurement cell developed with the Monte-Carlo N-Particle transport (MCNP) code. The calibration coefficient of 72 different matrices, representative of the waste produced at La Hague, were simulated and a statistical model using a multilinear regression was then established. This simulated-based approach provides a more robust and comprehensive estimation of the residual fissile mass without impacting accuracy thanks to the flux monitors and the transmission signals.

Material attractiveness of irradiated fuel salts from the Seaborg Compact Molten Salt Reactor

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

Using Lithium-6 Filter for Study of Dose Distribution with Max. and Min. Displacement of Prostate Inside the Body by BNCT Method

Physical dose assessment of prostate tumor and adjacent healthy tissues in ORNL, adult male (AM) and KTMAN-2 phantoms was performed using the Monte Carlo transport code. Two different spectra of MIT reactor (without and with lithium filter) were considered. To evaluate the effectiveness of this method in destroying cancer cells for different depths of the prostate, the prostates were moved three times with a distance of 3 cm towards the surface of the body. In addition, the effects of using a lithium filter with different thicknesses on the physical doses of the prostate and adjacent healthy tissues were investigated. It has been observed that decreasing the depth of the prostate increases the physical dose received by the tumor. When the prostate is placed at the maximum depth inside the body, the tissues located in the path of the beam receive the maximum amount of dose. However, the use of a lithium filter reduces damage to healthy tissues, so that when the prostate is near the surface of the body, the lithium filter increases the amount of dose delivered to tumors. In these conditions, the average energy of the penetration distance of epithermal neutrons and neutrons in tissues increases. By changing the location of the prostate gland inside the body, the physical doses of normal tissues did not change, but it was observed that increasing the thickness of the lithium filter significantly reduced the dose of normal tissues.

A method to predict space radiation biological effectiveness for non-cancer effects following intense Solar Particle Events

In addition to the continuous exposure to cosmic rays, astronauts in space are occasionally exposed to Solar Particle Events (SPE), which involve less energetic particles but can deliver much higher doses. The latter can exceed several Gy in a few hours for the most intense SPEs, for which non-stochastic effects are thus a major concern. To identify adequate shielding conditions that would allow respecting the dose limits established by the various space agencies, the absorbed dose in the considered organ/tissue must be multiplied by the corresponding Relative Biological Effectiveness (RBE), which is a complex quantity depending on several factors including particle type and energy, considered biological effect, level of effect (and thus absorbed dose), etc.While in several studies only the particle-type dependence of RBE is taken into account, in this work we developed and applied a new approach where, thanks to an interface between the FLUKA Monte Carlo transport code and the BIANCA biophysical model, the RBE dependence on particle energy and absorbed dose was also considered. Furthermore, we included in the considered SPE spectra primary particles heavier than protons, which in many studies are neglected. This approach was then applied to the October 2003 SPE (the most intense SPE of solar cycle 23, also known as “Halloween event”) and the January 2005 event, which was characterized by a lower fluence but a harder spectrum, i.e., with higher-energy particles. The calculation outcomes were then discussed and compared with the current dose limits established for skin and blood forming organs in case of 30-days missions.This work showed that the BIANCA model, if interfaced to a radiation transport code, can be used to calculate the RBE values associated to Solar Particle Events. More generally, this work emphasizes the importance of taking into account the RBE dependence on particle energy and dose when calculating equivalent doses.

Nuclear analyses for the integration of ITER equatorial Port 2

The present work is devoted to nuclear analyses in support of the ITER diagnostic Equatorial Port 2 (EP#2) integration. ITER EP #2 is a diagnostic port based on the long-modular Diagnostic Shielding Module (DSM) housing the following systems: Disruption Mitigation System (DMS) in DSM#1 and #3 and X-Ray Crystal Spectroscopy Core (XRCS-Core) in DSM#2. Ensuring adequate radiation shielding is a major challenge since the diagnostic systems require several apertures from the Vacuum Vessel (VV) through the Port Interspace (PI) and up to the Port Cell (PC). In the present study, a three-dimensional MCNP model of EP#2 has been developed, starting from the latest design available from Preliminary Design Review stage (PDR), and successively integrated into the reference 40° ITER C-Model. Comprehensive nuclear analyses have been carried out employing the D1SUNED v3.1.4 code based on the MCNP Monte Carlo transport code. Relevant nuclear quantities during and at the end of plasma operations have been evaluated: i.e., neutron and gamma fluxes and energy spectra along the port from the Diagnostic First Wall (DFW) up to the Bio-Shield Plug (BP), nuclear heating, neutron damage, helium and tritium production, and shutdown dose rate. This analysis allowed the identification of potentially critical areas, and therefore, the implementation of additional shielding options aimed at reducing the neutron streaming and mitigating of the radiation field in the PI region. In this work, the results of the analyses are presented and discussed. Some solutions to mitigate nuclear loads and to improve the shielding in PI area are proposed and their impact has been assessed. Finally, some recommendations for the optimization of the design of EP#2 are provided as well.

A new Excel-based Monte Carlo transport code for simulating energy response spectra and efficiencies in gamma-ray detection materials

This study presents and assesses a new, user-friendly, and Excel-based Monte Carlo photon transport code, specifically designed to simulate Gaussian energy broadened response spectra of gamma-ray detector materials. This vectorized code utilizes the EPICS2017 photoatomic data library, with unionized energy grid construction, to facilitate precise and efficient photon transport simulations in any user-defined materials composition. Performance benchmarks were conducted using PHITS 3.31, with both basic and detailed models of a gamma-ray scintillator system, and detection crystals of NaI(Tl) and LaBr3(Ce). These benchmarks spanned a wide array of photon source monoenergetic emissions. Detector model efficiency was determined through automated peak area analysis, employing algorithms from the Ortec GammaVision software. The results indicated only minor spectral differences between the Excel-based code and the PHITS models, particularly within the range of practical gamma-ray energies and near-zero source-detector distances. Efficiency curves demonstrated good agreement between the Excel-based code and PHITS models, especially within the energy range of approximately 200 keV to a few MeV, with increasing agreement as detector crystal size increased. Comparisons were conducted with the existing data in literature at source-detector distances of 2 cm and 5 cm, using a detector crystal size of 2”. These comparisons support the code’s validity and its implementation of a basic geometry. This new Excel-based code appears as a promising tool with potential applications in radiation detector characterization and materials research.

Add 3D Plotting to Your Monte Carlo Code In A Few Hours

Any Monte Carlo transport code can produce 2D slice plots of geometric inputs using sampling of cell or material identifiers at each pixels’ point in space; ray-traced 3D plots of simulation geometries are generally less available or require paid or closed-source software. We demonstrate the ease of implementing minimalistic ray-traced visualization algorithms using only the geometric querying operations found in any Monte Carlo transport code.

High-fidelity models for the online control of power ramps in fuel displacement systems – Revisiting the ISABELLE experiment in the OSIRIS Material Testing Reactor

The Jules Horowitz Reactor (JHR) is an MTR under construction at the CEA Cadarache, France. Its design and future operation build upon the lessons learned from the OSIRIS MTR. The CEA intends to transfer the knowledge accumulated with the ISABELLE1 loop of OSIRIS to the ADELINE loop of the JHR, both dedicated to power-ramp-type irradiation experiments, the purpose of which is to test the resistance of PWR fuel rod cladding under extreme levels of stress. In this article, we revisit the ETALISA experiment, a heat balance measurement experiment performed in 1992 for calibrating power ramps in ISABELLE1. We use high-fidelity modelling and simulation tools, especially the neutron-gamma TRIPOLI-4® Monte Carlo transport code, to calculate the detailed components of the heat balance, correction terms, and uncertainties. Comparisons between the simulations and the experiments show a very good agreement in the total linear heat generation rate of 400 W/cm at high power. The computed 2σ uncertainty is found to be 5%, a value essentially identical to the estimate derived in 1993 from an engineering approach. The use of modern simulation tools does not make it possible to improve upon this value, but provides a better understanding of the various components and corrections introduced in the total heat balance. The main limitations come from the ISABELLE1 online instrumentation, thermocouples and self-power neutron detectors, which set a limit on our very knowledge of the actual power ramp experimental conditions.

The effect of burnable absorbers on criticality and reactivity coefficient of VVER-1000 assembly

Burnable absorbers play an important role in nuclear reactor safety and exploring their effect on the reactor core behavior is an important issue in the design and operation of the reactor core. The aim of the present work was to discover the effect of burnable absorbers on criticality and reactivity coefficient of VVER-1000 assembly. The calculations were conducted with a 3-D Monte Carlo transport code MCNP6 and ENDF/B-VII.1 nuclear library. The Gd2O3 content varying from 0 wt.% to 8 wt.% was considered to complete the inter-comparison analysis between the criticality and the reactivity coefficients for three UO2+Gd2O3 fuel configurations. At the beginning of the cycle, there is a significant difference between the criticality (kinf) of assembly with and without Gd2O3, however, at the middle of the cycle those differences become very small and almost the same at the end of the cycle. The Doppler temperature coefficient values are always sufficiently negative and demonstrate a more negative trend with increasing gadolinium concentrations and fuel burnup. At beginning of the cycle, the moderator temperature coefficient value increases negatively as gadolinium concentration increases but, at middle of the cycle this trend does not occur. The fuel composition is predicted to be the reason behind this situation. At the end of cycle, there is no clear trend in the moderator temperature coefficient values with respect to Gd2O3 concentration. The absorbing effect of Gd2O3 appears to have diminished significantly. Overall, this research provides insights into the influence of the burnable absorbers on the neutronic parameters of the VVER-1000 assembly and its contribution to reactor safety.

Depletion Calculations for an Integral Small Molten Salt Reactor With Serpent

Abstract The molten salt reactor (MSR) concept is among the Generation IV designs considered feasible for providing clean, safe, sustainable, and economical energy supplies to the world's population. The depletion of fuel for a small modular fluoride molten salt reactor (sm-FMSR) with a closed fuel cycle based on the integral molten salt reactor concept has been investigated using Serpent. The Monte Carlo transport code Serpent has burnup capability and flow features that can be used to model fuel circulation and online fuel addition in an MSR. Three fueling schemes to control Serpent depletion cycles have been simulated and compared: step fueling (SF), continuous fueling with all fission products (FPs) accumulating in the reactor system (CFA), and continuous fueling with insoluble FPs separated from fuel (CFS). CFA and CFS require fewer depletion cycles that are longer in duration than the cycles required by SF, in order to maintain the effective multiplication factor (keff) within a working range over the seven years of the reactor fuel cycle. sm-FMSRs with SF and with CFA require similar quantities of “top-up” fuel, consume similar fuel (fissile) amounts, and result in similar fuel isotopic concentrations if keff is kept within a similar range. However, with separation of insoluble FPs from the circulating fuel, CFS gains a large reactivity worth due to the removal of FP poisons. This allows for reduction of fuel enrichment in both initial and total top-up fuel and leads to savings of a considerable fissile quantity in fueling MSR and in spent fuel. The Serpent depletion calculations require manual arithmetic calculations for adjustment of the Serpent built-in settings before the start of every calculation cycle for all three fueling schemes. Implementation of additional Serpent flow features in changing material volumes and flow constants would facilitate the simulation of the fuel depletion process and allow for more realistic simulations of fuel circulation.

Improvement of neutron sensitivity for lithium formate EPR dosemeters: a Monte Carlo analysis.

This work presents the computational analysis of the sensitivity improvements that could be achieved in lithium formate monohydrate (LFM) electron paramagnetic resonance (EPR) dosemeters exposed to neutron beams. Monte Carlo (MC) simulations were performed on LFM pellets exposed to neutron beams with different energy spectra at various depths inside a water phantom. Various computations were carried out by considering different enrichments of 6Li inside the LFM matrix as well as addition of different amounts of gadolinium oxide inside the pellet blend. The energy released per unit mass was calculated with the aim of predicting the increase in dose achievable by the addition of sensitizers inside the pellets. As expected, a larger amount of 6Li induces an increase of energy released because of the charged secondary particles (i.e. 3H ions and α-particles) produced after neutron capture. For small depths in water phantom and low-energy neutron spectra the dose increase due to 6Li enrichment is high (more than three orders of magnitude with respect to the case of with 7Li). In case of epithermal neutron beams the energy released in 6Li-enriched LFM compound is smaller but larger than in the case of fast neutron beams. On the other hand, the computational analysis evidenced that gadolinium is less effective than 6Li in improving neutron sensitivity of the LFM pellets. Discussion based on the features of MC transport code is provided. This result suggests that 6Li enrichment of LFM dosemeters would be more effective for neutron sensitivity improvement and these EPR dosemeters could be tested for dosimetric applications in Neutron Capture Therapy.

Validation of MCkeff code using ICSBEP results

Validation of the indigenously developed Monte Carlo program, MCkeff, is carried out by comparing the results of International Criticality Safety Benchmark Evaluation Project (ICSBEP) problems with that of standard legacy code MCNP (general-purpose Monte-Carlo N-particle transport code). Code-to-code criticality validation is essential before the usage of MCkeff to model various kinds of fissile systems. The Evaluated Nuclear Data Files (ENDF/B VII.I in ACE Format) used in MCNP is expended in the present study, thereby avoiding the differences in the results due to cross-section data. Thus, the deviations in the computed results will reflect only the differences arising out of differences in algorithms employed in these codes. Plutonium metal benchmark problems of different geometrical configurations were selected for validation purposes. All problems were simulated using MCkeff with the same inputs as in MCNP validation, and the computed results of keff (neutron multiplication factor) of various systems were compared. The results from MCkeff agree with that of MCNP, and the maximum deviation between them is less than 0.1%. Hence, the code MCkeff can be used to analyze other metal systems problems. Further validation work on the different fissile systems with other physical forms is in progress.

Activation Analyses of Disposal Options for Irradiated Be12Ti

The activity and disposal options for irradiated Be12Ti were assessed for the HCPB DEMO blankets making use of a code system that enables performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. The dedicated full-scale geometry MCNP model of the 11.25 degree HCPB DEMO torus was adapted to the requirements for the coupled 3D neutron transport and activation calculations. Special attention was paid to the use in the activation calculations of the commercial materials containing technological impurities. This has a crucial effect on the results and the impurities must be accounted for in any nuclear safety analyses. The short-term activity is formed by the radionuclides produced through the activation of Be and Ti nuclei and the long-term activity is formed by the products of the neutron irradiation taking place on the impurities. A prerequisite for the disposal or the recycling of the irradiated Be12Ti is its deep detritiation; otherwise, the very high-tritium activity would fully prevent any attempt for its treatment. The most preferable is the use of the Be12Ti with the composition including less material impurities, especially uranium. There could be the option to dispose the Be12Ti intermediate-level wastes in the French repository after 1 year of cooling, assuming the detailed control of the impurities that fulfil the French authority requirements. The USA near-ground repositories could be an alternative to the European sites. The recycling of the irradiated Be12Ti must first be elaborated and approved to ensure its treatment in a safe and efficient way. The remote handling technique must be developed for the re-fabrication of the Be12Ti blocks.

Benchmarking of Proton- and Neutron-Induced Nuclear Data with MCNP6.2 Using SINBAD Experiments

The effects of both proton and neutron nuclear data libraries on the neutron field produced in light targets and transported through shielding material were assessed. The general-purpose nuclear data libraries JENDL-4.0/HE, TENDL-2017, ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.1.2, and JEFF-3.3 were benchmarked using the MCNP6.2 Monte Carlo transport code. Three experiments from the SINBAD database, namely, 52-MeV protons hitting a carbon target and 43- and 68-MeV protons hitting a 7Li target, were selected for benchmarking. This selection was made according to their relevance for the MYRRHA accelerator-driven system to support the design of beam dumps and shielding of the 100-MeV section of the MYRRHA accelerator. It is demonstrated that the prevailing factor determining the transmitted neutron spectrum shape is the proton library from which the secondary neutrons are sampled. The choice of neutron library applied for the secondary neutron transport in the carbon and lithium targets, concrete and iron shielding, is of second-order significance.

Operational radiation protection challenges for the LHC experiments.

Radiation protection physicists at CERN are often required to assess residual activation for the Large Hadron Collider (LHC) experiments during stop periods in order to ensure adequate optimization during planned exposure situations and to establish proper procedures for the radiological control of materials. Given the complexity of the facilities and of the high-energy and mixed fields inducing the activation, Monte Carlo transport codes are an essential tool to simulate both prompt and residual radiation. The present work highlights the challenges of assessing residual dose rates for the LHC experiments in shutdown configurations and of establishing residual activation zonings. For the latter, a method based on fluence conversion coefficients was developed and is efficiently employed. The practical example of the assessment of the activation of the 600 tons of austenitic stainless steel in the future Compact Muon Solenoid (CMS) High Granularity Calorimeter will be used to demonstrate how these challenges are dealt with and the capabilities of the method developed.