© The Author(s) 2025. Published by EDP Sciences and China Science Publishing & Media Ltd.

1 Introduction

All nuclear facilities need to be decommissioned after their permanent shutdown. Differing from nuclear reactors in operation, nuclear decommissioning usually involves continuously changing radioactive environments [1]. The radiological hazards encountered during decommissioning will also vary from those experienced during the operational phase. Generally, the average collective dose of nuclear power plants (NPPs) in cold shutdown/under decommissioning is lower than during active operation [2]. However, radiation release during decommissioning is more difficult to predict and control, occurring within a much shorter time frame, potentially spanning days or months. Moreover, considerable variation exists between different reactors and over time, and previous decommissioning efforts may offer limited guidance for upcoming projects especially different kinds of experimental reactors. Therefore, the design of effective radiation protection systems is crucial to minimizing the potential radiation exposure to on-site workers before construction begins. There radiation protection measures include establishing radiation protection zones, optimizing internal and external radiation protections, radiation monitoring protocols, as well as improving ventilation systems. Through these measures, radiation risks can be effectively managed, ensuring the safety of both workers and the public.

Several guidelines have been developed to address radiation protection during nuclear decommissioning. The International Atomic Energy Agency (IAEA) has issued Tecdoc-1954, providing comprehensive guidance on radiation protection in decommissioning from the perspectives of management, planning, and execution [1]. Germany has introduced the IWRS II guidelines to address radiation protection for personnel involved in maintenance, repair, waste management, and dismantling activities at nuclear facilities [3, 4]. Research has also examined the impact of protective clothing, such as air suits, on exposure time, suggesting that exposure time can be extended by up to 65% depending on the type of work and protective equipment [5].

In addition to conventional radiation protection measures, visualizing the radiation field has been shown to enhance radiation protection efforts [6, 7]. Digital twin technology, an emerging field that combines time-evolving physics-based and data-driven models [8], allows for more vivid simulation of radiation fields and the entire decommissioning process. Based on the digital twins, it becomes more efficient to conduct employee training, simulate radiation fields, optimize decommissioning processes, and improve radiation protection, etc. 3D simulation performed by Halden Planner and VRdose software enables real-time estimation of risks to workers, facilitating identifying optimal worker routes, shielding configuration, order of subtasks, and so on [9].

The Tsinghua University Shielded Experimental Reactor (TUSER, in Beijing), the first independently designed research reactor in China, is being prepared for its decommissioning. The decommissioning requires comprehensive radiation protection planning. However, there is still a lack of experience in decommissioning nuclear facilities in China, especially in terms of radiation protection. While occupational radiation protection experience in the context of decommissioning exists, expertise in this specialized area remains scarce among many stakeholders. This study investigates the safety measures implemented in the decommissioning of the TUSER, with a focus on the design of radiation control zones, the establishment of secure personnel pathways, and the integration of internal and external radiation protection and monitoring systems. Additionally, a comprehensive ventilation system and digital twin system were developed to ensure safe working conditions.

2 Introduction of TUSER

The TUSER began construction in 1960 and reached criticality for the first time in 1964. It was initially designed and constructed for the operation of power levels up to 2 MW. However, it was upgraded to 4.8 MW in 1975. It was permanently shut-down in 2009 after an operational period of 40 years without any incident, and accomplished many experimental tasks, using low-enriched U (10%). Now, it has started the decommissioning project from this region, implementing the strategy of immediate dismantling and decommissioning. The operations are conducted by the Reactor Decommissioning Department. The transportation and temporary storage of spent fuel have been completed in 2015. Table 1 gives the history and characteristics of the TUSER.

3 Personnel dosimetry and free-release limits

The target dose management objectives for on-site personnel and the public during decommissioning are outlined in the Basic Standards for Protection Against Ionizing Radiation and for the Safety of Radiation Sources (GB 18871-2002) and the radiation protection program of the Institute of Nuclear and New Energy Technology (INET). These objectives stipulate that the annual dose for on-site personnel should be less than 4.5 mSv, while that for the public should remain below 0.01 mSv.

The limits for wastewater discharge set by INET specify that the α activity concentration in wastewater must be less than 1 Bq/L, while the β activity concentration must not exceed 3.7 Bq/L. The limit for surface radioactive residue levels are determined as one-fiftieth of that for radioactive surface contamination in the workplace according to the GB 18871-2002. The α activity concentration is restricted to less than 0.08 Bq · cm⁻², and the β activity concentration should be less than 0.8 Bq · cm⁻². For radioactive packages, the limits are based on the Regulations for the Safe Transport of Radioactive Material (GB 11806-2019). For the surface contamination level, the β and γ emitters, as well as low-toxicity α emitters should be less than 4 Bq·cm⁻², all other α emitters is less than 0.4 Bq · cm⁻². The highest radiation level at any point on the external surface of the package or steel box is less than 2 mSv/h. The residual surface contamination levels of equipment and tools for direct reuse refer to the limits specified in the GB 18871-2002. The total α activity concentration is less than 0.4 Bq · cm⁻², and the β activity concentration should be less than 4 Bq · cm⁻².

According to the radiological characterization of TUSER, the remaining radionuclides are mainly ¹⁴C, ⁵⁵Fe, ⁶⁰Co, ⁶³Ni, ⁹⁰Sr, ¹³⁷Cs, and ¹⁵²Eu. The limits for specific activity of these key radionuclides are listed as follows: for recycling and reuse of scrap steel and aluminum, the free-release level of solid waste, extra-low radioactive waste, and remaining radioactive contamination level of the soil. All key radionuclides are presented in Table 2.

4 Radiological protection

4.1 Radiation control zoning and routes for workers

Radiation control zoning and designated worker routes are essential measures during decommissioning. These zones are determined based on the degree of contamination within different areas. On-site activities such as equipment maintenance, repair, and other tasks typically do not involve high-level radioactive environments under normal conditions. Based on the specific characteristics of the TUSER ventilation system, the operational area is divided into controlled and supervised zones. A schematic diagram of the controlled and supervised zones on the first floor is provided (Fig. 1).

Fig. 1 Radiation control zones of the 1st floor (the controlled zones are marked as red, while the supervised zones are blue).

The controlled zones include the reactor hall from the first to the fourth floor, the temporary waste storage area, decontamination room, cutting room, analysis laboratory, hot cell area, pump room, and more. The supervised zones mainly consist of the male and female bathrooms on the first floor. Entry into the controlled zone is not permitted without authorization.

The process for workers entering and exiting the controlled zone during decommissioning is as follows (Fig. 2). First, in the changing room, workers change from their daily clothes into work clothes, wear work shoes and protective gear. After putting on a direct-reading personal dosimeter and a photoluminescent personal dosimeter, they enter the site to carry out their tasks. After completing the on-site work, workers undergo surface contamination monitoring at the exit of the control area. If the radioactive contamination is detected during the monitoring, the protective clothing, masks, gloves, shoe covers, etc., are collected separately, and the workers undergo decontamination. After decontamination, the hands, feet, and body surface of the workers are measured again. Only after passing the measurement can the workers enter the changing room to wear their daily clothes, while registering the time of entering and exiting the controlled area, the operations, and other information. If no contamination is detected during monitoring, the protective clothing, masks, gloves, shoe covers, etc., are collected, and the workers change into their daily clothes, while also registering the time of entering and exiting the controlled area, the operations, and other information.

Fig. 2 Illustration of staff entering and exiting the controlled zone.

4.2 Internal/external radiation protection and monitoring

4.2.1 Internal/external radiation protection

All radiological hazards to which workers are exposed during decommissioning activities are to be considered and continuously evaluated. General protection measures include the remote-controlled operations, fixed hoisting routes, and packaging high-radiation-level waste with limited-positioning baskets and steel boxes for transport. The radioactive waste is stored in designated areas, and dedicated workers flow channels are established. In the decommissioning workplace, ventilation systems or local air tents are set up. Before workers enter the construction site, the ventilation system must be activated to reduce the aerosol concentration in the workplace and minimize the inhalation of radioactive gases by operators.

Personal internal radiation protection for operators entering the controlled area is implemented through a tiered protection system, classified as: Level I, Level II, and Level III. Level I is the highest level which involves wearing forced-air respirators, encapsulated suits, safety helmets, rain boots, foot covers, fine mesh gloves, industrial latex gloves, plastic sleeve covers, coarse mesh gloves, and anti-contamination temporary plastic products. Level II/III protection compared to Level I are listed in details in Table 3. In the event of serious abnormal incidents during decommissioning, after evaluation by experts, the management and control of workers at the construction site are required to follow Level I requirements. During stages such as the dismantling of components inside the reactor, removal of the biological shield, decommissioning of redundant systems, and dismantling of remaining facilities, as well as during offline decontamination, the management and control of workers entering the construction site follow Level II requirements. For other construction activities within the controlled area, the management and control of workers entering the construction site follow Level III requirements.

Some experiments are conducted to characterize and control the release of radioactive aerosols. During the pre-decommissioning phase, radioactivity and health effects of aerosols released during the cutting, drilling and components transfer were evaluated. The gross α and β radioactivity in most scenarios was compared to the background level, while the plasma arc cutting had a maximum α and β radioactivity over 0.10 and 0.14 Bq/m³, respectively. The internal exposure dose caused by ¹³⁷Cs aerosols were evaluated, different AMADs (activity median aerodynamic diameter) can cause a fourfold discrepancy [10]. Therefore, during the decommissioning phase of the TUSER, measurements of the particle size distribution and radionuclide composition of aerosols produced during the graphite retrieval, plasma arc cutting and diamond wire cutting are conducted.

An assessment and verification of the aerosols produced by different processes and in different scenarios is performed. Simulants are created based on the size, material, and radionuclide composition of the reactor objects. For the graphite retrieval process, most graphite dust is in the micrometer size range (mass medium aerodynamic diameter, MMAD > 10 μm), which easily deposited or filtered by the overall/partial ventilation. For the cutting process, the metal simulant is cut using plasma arc inside a closed air tent. Sampling pipelines are deployed within the air tent, connecting to a scanning mobility particle sizer to determine the particle size distribution [11, 12]. Air samplers are used to collect aerosols to determine the radionuclide composition [10]. By adjusting plasma cutting power, cutting speed, and other parameters, information about the particle size distribution, concentration, and radionuclide composition of the generated aerosols is obtained through the two sampling measurement devices. Preliminary results show that the aerosols generated during the plasma arc cutting has two diameter peaks at less than 10 nm and ~200 nm, influenced by the cutting material, electric current, thickness, and so on. This provides a reference for optimizing cutting parameters and necessary parameters for the selection and setup of air purification devices and the optimization of workers protection measures.

4.2.2 Internal/external radiation monitoring

Radiation monitoring includes the monitoring of workplace and individuals, which is conducted throughout the decommissioning. The main aspects of workplace radiation monitoring include γ dose rate, aerosol concentration, surface contamination, radioactive residue of buildings and structures and radioactive level of waste. γ dose rate monitoring includes monitoring of γ dose rates during operational activities, within buildings, and in the surrounding area within the 20 m range. Aerosol concentration monitoring involves aerosol concentrations at the work sites inside buildings, in the surrounding area within a 20 m range, and at the downstream end of ventilation duct filters. Surface contamination monitoring includes monitoring the surfaces of objects and the body surfaces of workers. Radioactive residue monitoring of buildings and structures involves on-site analysis using portable γ spectrometers and laboratory analysis of samples taken. Radioactive level monitoring of waste is to mainly identify the types and activities of nuclides contained in waste in steel drums and steel boxes.

Monitoring of individuals engaged in on-site operations include the assessment of external exposures and its internal exposures with special emphasis on alpha particles. Workers entering the controlled area for on-site operations also wear thermoluminescence dosimeters (TLDs) and direct-reading personal dosimeters, the latter of which is for real-time monitoring of personal external radiation dose. During the project implementation period, internal exposure monitoring for personnel primarily focused on airborne aerosol surveillance. Air samples from operational areas were collected and analyzed to quantify aerosol concentrations. The internal exposure doses to workers were then estimated by integrating the measured aerosol data with on-site working conditions and task-specific parameters. For operations involving high doses, measures are taken to control the operation time of the workers or to rotate tasks to distribute the dose, ensuring that the individual exposure dose is kept within the target values.

4.3 Ventilation system

4.3.1 The overall ventilation system

The airflow organization principle of the ventilation system is that air flows from clean areas to contaminated areas. Within contaminated areas, the air flows from regions with low radioactivity to those with high radioactivity. The radioactive areas are isolated from the outdoor environment with sealed doors and windows, maintaining a negative pressure relative to the atmosphere. The clean areas are separated from the outdoor environment with ordinary doors and windows, maintaining atmospheric pressure balance between indoors and outdoors.

Systems that may generate radioactive aerosols have their exhaust air passed through pre-filters and high-efficiency particulate air filters (HEPA filters, the filtration efficiency for 0.3-μm particulates is higher than 99.9999%) before being discharged into the atmosphere through a 60-meter-high chimney. Pressure differential gauges are installed before and after the filters to monitor the filter resistance status. All HEPA filters are subject to on-site efficiency testing and leak detection using a sodium flame method after installation.

There are six sets of process exhaust systems (Fig. 3): I system for drying boxes, storage pool, material channels, and fast neutron channels; II system for hot cells and radiation loop compartments; III system for the upper space of the reactor pool and broken element boxes; IV system for storage pool rooms, thermal loops, radiation loop laboratories, material channel anterooms, pump rooms, and filter rooms; V system for deactivation fume hoods; VI system for the reactor hall. The general exhaust system is mainly used for non-radioactive work areas, such as toilets, corridors, dressing rooms, electrical rooms, and ordinary physics and chemistry laboratories. These exhaust systems generally use an air exchange rate of 3–5 times per hour to maintain fresh air in these areas and meet temperature and humidity requirements. For the graphite retrieval process, the overall ventilation can achieve a filtration efficiency higher than 90%.

Fig. 3 Ventilation system of the experimental shielding reactor.

4.3.2 The partial ventilation system

The internal components of the reactor are dismantled in air. Unlike underwater dismantling, the dismantling in air may release lots of aerosols. The remote control is applied to most instruments to minimize on-site operations by workers. To control the spread of aerosols during the dismantling process, and an air tent with a local exhaust and purification system is added above the reactor pool. Aerosol suppression technology is also employed to control aerosol diffusion, such as water flow or spray droplets. During the cutting of components inside the reactor, a local suction method is used to promptly draw the aerosols to the purification system, preventing the deposition of aerosols, which could then fall during subsequent hoisting operations.

The biological shield is cut layer by layer from top to bottom using a drilling machine and a diamond wire saw machine. The base of the biological shield is dismantled by a robot with a matching removal tool head. During the cutting process, the hall exhaust is activated, and techniques such as spraying and other aerosol suppression methods are used to control aerosol diffusion during the dismantling process. Other works like graphite blocks retrieval, plasma torch cutting or decontamination, are all conducted with partial ventilation system. In some scenarios, a mobile vacuum cleaner is applied to collect dust in time.

4.4 The digital-twin system

Before initiating decommissioning construction, a digital-twin system was established and validated to smooth the decommissioning process and ensure the health and safety of on-site operators (Fig. 4). This digital-twin system leverages development platforms including Visual Studio.NET (C/S system) and Java (B/S system), along with the independently developed 3D engine RINVR. Complementary software such as CATIA, Maya, 3ds Max, as well as Photoshop are utilized for 3D modeling and material mapping. Using construction drawings and 3D laser scanning data of the TUESR, a 3D model and corresponding scenarios are constructed. A validation platform is developed to analyze scenarios and assess dose rate, enabling the evaluation of decommissioning plans. Additionally, a management platform is created to facilitate decommissioning activities, including data storage, plan formulation, progress tracking, and archival.

Fig. 4 The structure of the digital-twin systems.

The radiation field simulation module utilizes data from radiological characterization to calculate the radiation dose field (Fig. 5). By integrating these calculations with the 3D model, the spatial distribution of the radiation dose field is visualized. The module incorporates both rapid and precise calculation algorithms. The rapid calculation algorithm applies the point kernel integration method, which is an empirical calculation technique that is not constrained by the spatial geometry of the field or the shielding thickness of the radiation source [13]. The rapid calculation algorithm can evaluate various decommissioning plans in a short time, identifying the optimal solution. Although the fast mode significantly reduces computation time, it introduces some degree of error. While the precise calculation algorithm performs detailed three-dimensional radiation field calculations for the selected plan. The Monte Carlo method is used for the precise calculation, which is well-established with relatively low convergence rate [14]. This ensures the provision of quantitative radiation dose data to inform design workers.

Fig. 5 Flowchart of radiation field simulation for decommissioned nuclear reactors based on digital twins.

During the decommissioning process, operators involved in dismantling and cutting are designated as dose monitoring points. The system provides real-time monitoring of individual and cumulative radiation doses, which are displayed in Figure 6. The human dose module offers functionality to set dose warning thresholds, adjust sampling intervals, select statistical methods (e.g., roaming or specified paths), configure task movement speed in specified path mode, and visualize dose values and cumulative dose charts in real-time. As the decommissioning activities proceed, certain input parameters, as illustrated in Figure 6, are subject to variation. Therefore, it is necessary to modify these parameters according to measurement results and on-site situations, and then recalculate the radiation field and radiation doses. Combined with VR (virtual reality) technology, the digital twin has the potential to serve as an effective training instrument, enabling personnel to rehearse on-site operations, enhance operational efficiency, and curtail radiation exposure.

Fig. 6 Dose assessment results of an avatar (inst. and cum. are the abbreviations of the instantaneous and cumulative, respectively). (a) An illustration of the avatar inside the TUSER; (b) The cumulative and instantaneous effective dose received by the avatar during its 288-second movement trajectory; (c) The instantaneous and (d) cumulative equivalent dose received by the brain and heart of the avatar.

5 Conclusion

In conclusion, effective radiation protection is paramount during the nuclear decommissioning process to ensure the safety of both workers and the public, as well as to guarantee the smooth and successful execution of decommissioning projects. Achieving this requires thorough and proactive planning, with comprehensive radiation protection strategies implemented well in advance to minimize unnecessary exposure and risks. To further optimize safety and radiation control, it is critical to establish and adhere to stringent standards and best practices. These should focus on limiting radiation exposures, integrating advanced protection technologies, and ensuring their effective implementation in the workplace.

The incorporation of innovative technologies, such as the digital twin system, plays a pivotal role in improving radiation control zoning, validating decommissioning processes, and optimizing overall decommissioning plans. This system enhances the efficiency of verification procedures and facilitates data-driven decision-making by providing real-time monitoring and simulation capabilities. While external radiation exposure is often more easily managed and presents immediate risks due to its higher radiation doses, internal exposure remains a hidden hazard, underscoring the necessity for continuous, real-time monitoring and robust warning systems. The use of 3D modeling, radiation field simulation, and precise dose assessment modules within the digital twin system ensures the accurate modeling of radiation distributions, enabling informed decision-making and optimizing the decommissioning workflow.

Funding

This work was supported by LingChuang Research Project of China National Nuclear Corporation.

Conflicts of interest

The authors have nothing to disclose.

Data availability statement

This article has no associated data generated.

Author contribution statement

Xiaotong Chen: Conceptualization, data curation, formal analysis, investigation, methodology, validation, visualization, writing – original draft, writing – review & editing.

Zhenzhong Zhang: Conceptualization, funding acquisition, project administration, resources, supervision, validation, writing – review & editing.

Haisheng Liu: Data curation, project administration, validation, writing – review & editing.

References

All Tables

All Figures

	Fig. 1 Radiation control zones of the 1st floor (the controlled zones are marked as red, while the supervised zones are blue).
In the text

	Fig. 2 Illustration of staff entering and exiting the controlled zone.
In the text

	Fig. 3 Ventilation system of the experimental shielding reactor.
In the text

	Fig. 4 The structure of the digital-twin systems.
In the text

	Fig. 5 Flowchart of radiation field simulation for decommissioned nuclear reactors based on digital twins.
In the text

Fig. 6 Dose assessment results of an avatar (inst. and cum. are the abbreviations of the instantaneous and cumulative, respectively). (a) An illustration of the avatar inside the TUSER; (b) The cumulative and instantaneous effective dose received by the avatar during its 288-second movement trajectory; (c) The instantaneous and (d) cumulative equivalent dose received by the brain and heart of the avatar.

In the text