Low-level Waste Research Articles

Uranium quantification in radioactive waste drums using In-Situ Object Counting System (ISOCS) for safeguards purposes

For the peaceful use of nuclear energy, materials, and technology, various methods have been developed for safeguards purposes. Among them, the In-Situ Object Counting System has demonstrated its usefulness and reliability in various nuclear material verification activities, due to its capability of operating without standard materials. In the present study, the performance of In-Situ Object Counting System for uranium quantification was evaluated using an experimental drum at KEPCO NF as well as low-level radioactive waste drums at the Korea Radioactive Waste Agency. The results showed that In-Situ Object Counting System could quantify uranium in radioactive waste drums, with a relative error ranging from −1 to −67 % compared with destructive assay results. This study also examined, using the Monte Carlo N-particle transport 6.2 code, the factors that affect uncertainty in uranium quantification: geometrical conditions, spectrum quality, heterogeneity, and the self-shielding effect. These findings could prove useful in efforts to optimize the In-Situ Object Counting System as a tool for nuclear material verification in the safeguards context.

Synthesis and Evaluation of the Extraction Efficiency of Pristine Zeolite Na-A to Remove ReO4- Ions (Surrogate of 99TcO4-) from a Simulated Low-Level Waste Solution.

Zeolite Na-A was synthesized through hydrothermal, alkali-fusion, and sonochemical methods, using kaolin as an economically viable precursor. The synthesized zeolite Na-A samples were characterized using XRD, FT-IR, SEM, and Raman spectroscopy. Specific surface area and pore size distribution analyses (by BJH and DFT models) were conducted using a BET surface area analyzer. Additionally, thermal degradation studies were performed by TG-DTA to check the thermal stability of zeolite Na-A at high temperatures. Furthermore, kaolin-derived zeolite Na-A was employed for the extraction of perrhenate ions (ReO4-), which are a nonradioactive surrogate for pertechnetate ions (99TcO4-) without any postsynthetic modifications (using toxic surfactants, etc.) utilizing in situ modification of the solution medium for the first time. The sonochemically synthesized zeolite Na-A demonstrated superior sorption performance for the solid-phase extraction of ReO4- ions from the simulated low-level waste solution. The adsorption process was found to follow pseudo-second-order kinetics. The Langmuir isotherm model fit perfectly with the experimental data (R2 = 0.997) and exhibited a maximum sorption capacity of 926.8 mg/g at pH ∼11, showing superior sorption capacities compared to those of the numerous materials reported earlier. XPS confirmed the speciation of extracted rhenium as NH4Re(VII)O4, providing critical insights into the adsorption mechanisms and validating the suitability of the sonochemically synthesized zeolites toward ReO4- sorption. Furthermore, Raman studies of ReO4- adsorbed zeolite Na-A reflect the absence of characteristic breathing bands, indicating the closure of the pore openings due to the occupancy of adsorbate moieties within the pores. This study not only highlights the utilization of sonochemically synthesized zeolite Na-A as an efficient, benign, and cost-effective adsorbent for 99TcO4- removal from nuclear waste but also emphasizes its potential sustainable applications in various other industrial processes such as wastewater treatment, catalysis, gas separation, pollution control, and resource recovery from industrial effluents and in the pharmaceutical industry for selective ion removal.

Sensitive imaging of actinide materials in shielded radioactive waste

This paper reports on the development of a method for enhanced non-destructive assay (NDA) of radioactive waste using the novel technique neutron-gamma emission tomography (NGET). The technique relies on the detection of correlated fast neutrons and gamma rays emitted in spontaneous or induced fission. It is based on fast organic scintillators and enables sensitive detection and three-dimensional (3D) localization of the fission events. The technique is passive and does not require moving components. In this work, we apply the NGET technique to the category of radioactive waste which is often referred to as historic or legacy waste. This can include mixed wastes encased in shielded containers many decades ago, before the advent of detailed waste description criteria. These low or intermediate level wastes are often associated with lacking, limited or conflicting documentation. This poses a challenge when assigning the waste to the proper disposal route as well as in deciding whether the waste needs to undergo sorting and conditioning to fulfil waste acceptance criteria both with regards to safe interim storage and to its ultimate disposal. Actinides, such as isotopes of uranium and plutonium, with their typically long half-lives and decay chains are of special interest in this regard since they may challenge the long-term safety assessment in repositories predicated on mainly shorter half-life radionuclides if undetected. Accurate identification and localisation of actinides is also important from a safeguards perspective, especially since they are generally difficult to detect and localise by established passive means due to their relatively weak radiation emissions, in particular in shielded containments and in the presence of strong radiation fields from other radioactive materials. In this paper we present findings of measurements on shielded containments of long-lived radioactive waste performed at the Studsvik site in Sweden, as well as measurements on a laboratory assembly simulating a grouted waste drum. Similarities and differences between the novel NGET technique and a commercially available gamma imaging system are also briefly discussed.

Potential amendments of coal fly ash-derived zeolite to beryllium contaminated soil at a legacy waste disposal site

Management of Be contamination using industrial solid waste or solid waste-derived amendments is not well understood. This study investigated the potential of Australian coal fly ash (CFA), derived synthesized zeolite (SynZ) and chitosan-modified zeolite (ModZ), for Be immobilization at the Little Forest Legacy Waste Site (LFLS), a low-level radioactive waste disposal site near Sydney, Australia. In laboratory simulation experiments, the SynZ and ModZ were separately applied as an amendment to both naturally contaminated soil and simulated contaminated (spiked) soil. Different techniques, including pore water (PW), batch desorption, and microbial activities were assessed to provide insight into immobilization mechanisms. Results revealed that amendment of 2% ModZ in soils, substantially decreased Be concentrations in PW (PWBe) ranging from 13.3% to 99.5% across all concentrations of Be. In contrast, PWBe increased while using SynZ, which could be attributed to the increased solubility of different organic-inorganic elements in PW. Moreover, batch desorption using Milli-Q water, simulated acid rainwater [H2SO4/HNO3 = 60/40, (v/v), and 0.11 M acetic acid solution also revealed similar patterns of Be immobilization as found in PWBe analysis. Soil amendments boosted microbial biomass carbon, and phosphorous (MBC,P), along with basal respiration (BRCO2). This indicates increased microbial activities, which are linked with environmental eco-friendliness. This effect was substantially noticed in ModZ-amended soils, exhibiting up to 22 times higher in BRCO2 values compared to unamended soil. Additionally, reduced PWBe was correlated with soluble organic-inorganic elements, desorbed Be in the batch study, and soil MBc. The differences in behavior between SynZ and ModZ underline the importance of carefully studying the various potential amendment materials and the need to evaluate their performance before application in field situations. This study highlights ModZ's effectiveness in eco-friendly Be immobilization, underlining the role of organic functional groups in zeolite architecture, a key factor in controlling Be in soils.

A mechanized assembly for erecting soil-cement barriers to protect agricultural lands from low-active waste during flood

The object of this study is the process of constructing a waterproof underground barrier from local soil using loess loam and medium-sized sand as an example. The values of waterproofing of soil cements on different types of soil and the norms of net labor consumption corresponding to them were obtained. The research solves the task of protecting agricultural lands in the areas of geological burials of low-level radioactive waste. Configuration of a mechanized drilling mixing assembly for installing a waterproof barrier in the field has been proposed. Waterproofness of soil cement on clay and sandy soils was determined experimentally. The corresponding norms of the net consumption of time for manufacturing the barrier in these soils have been determined. Units of the mechanized assembly are repairable, utilize widely available materials, parts, and mechanisms. As a result of studies on the labor intensity of the installation of soil cement, a net labor rate was established, which varies from 35 to 52 min/m3 depending on the type of soil. The corresponding values of waterproofness, which ranged from W6 to W14, were determined, which substantiates the possibility of effective functioning of soil-cement underground barriers. Options for the structural solution for impenetrable barriers have been offered. The values of the obtained net time rates are explained by the high degree of mechanization of the technological process and the use of local materials. A distinctive feature of the results is the emphasis on the data collection on the time consumption of machines and mechanisms, the characteristics of local soils under production conditions, and the use of local materials. The assembly implies using affordable and repairable units that can be serviced right in the field. Given this, the manufacturability of the proposed process was defined, confirmed by time-keeping studies. The realm of practical application of the reported results are sites within flat areas with sandy or loess bases. Barrier solutions are designed exclusively for geological waste storage facilities under the conditions of a high level of groundwater and the presence of a waterproof layer at the base

Low level radioactive waste treatment by coagulation flocculation technique: a review

Low level radioactive waste treatment by coagulation flocculation technique: a review

Leachability of low-level radioactive waste from waste disposal repository at Tarapur, India

Leachability of low-level radioactive waste from waste disposal repository at Tarapur, India

Comparing the short-term leaching behaviour of cements containing supplementary cementitious materials in different leachants for low-level waste encapsulation

Pulverized fly ash (PFA) is commonly used as a supplementary cementitious material (SCM) in cement and is permitted for use in cements and grouts used for the encapsulation of low-level nuclear waste. However, with a global decrease in availability of PFA predicted in the near future, alternative SCMs are required. In nuclear encapsulation, a key parameter for a new SCM is that it does not change the leaching behaviour of the grout. We present promising initial leaching data from two SCMs: metakaolin and waste paper fly ash. These were tested for 24 h in two leachants, deionized water and an artificial groundwater, tracking the pH and leachability of a range of elements as the samples equilibrated with the leachants. The study suggests that for many elements, both metakaolin and waste paper fly ash show a very similar leaching behaviour to PFA grout formulations despite differences in the pH equilibration. In general, the tests suggest that the composition of the elute is dominated by leaching from the cementitious phases, and not the SCM. This suggests that further longer-term leaching studies are required as part of testing for SCM suitability for nuclear grouts in order to understand the connections between leaching and grout microstructure.

Post closure safety assessment for near surface disposal

Near-surface facilities which are designed on or below the surface are typically used for Low Level Waste (LLW) and short-lived Intermediate Level Waste (ILW) disposal. Due to the site characteristics changes such as climate in the long-term period, a comprehensive disposal Safety Assessment (SA) is required to predict the future situation. The main aim of this paper is to make a safety assessment of a trench type near surface disposal facility in Iran based on ISAM methodology recommended by the International Atomic Energy Agency (IAEA). In such a manner, a surface water erosion, a shepherd, and a human intrusion are considered as design and alternative scenarios to model using AMBER code for a period of 100,000 years. Results presented; the maximum radiation is related to human intrusion scenario in which the total dose radiation reached to 0.3 mSv/y in the first year after the site passive institutional control. Moreover, the dose released from the investigated scenarios in the post-closure phase was less than 1 mSv/y, which is the IAEA regulatory body safety criteria. Comparison of the results showed that the current site is designed and constructed in accordance with safety guidelines which are established by the IAEA and Iran National Regulatory Authority (INRA).

Design and structural analysis of high payload C1 container

Abstract To meet the need of storing the nuclear facility dismantling wastes in Taiwan, National Atomic Research Institute (NARI) plans to develop a new type of low-level radioactive waste container based on the exterior specifications of the existing C1 container but with higher payload. Currently, the design of this container has been finished, and the associated structural analyses specified in the regulations have also been done. In this article, it summarizes what NARI has performed and what NARI plans to do for the development of this container. It includes not only the information of C1 container and its actual applications but also the design specifications and the associated numerical analyses of the container under development. Based on the analysis results, the container under development satisfies the requirements of the low-level radioactive waste storage container and the industrial package type 2. Currently, NARI has entrusted a local vendor to fabricate the prototypes of this container and perform the specified tests in the second half of 2024. The usage permit application of this container will be submitted to the authority agency in the first half of 2025. Hopefully, the usage permit could be granted in 2026.

Discerning β-emitting radioactivity from caesium-137 γ radiation via bremsstrahlung for in-situ groundwater monitoring

A qualitative approach to discern the relative contributions of 137Cs and 90Sr in-solution with a monitoring probe is described, based on a comparison of X- and γ-ray photon spectra. The bremsstrahlung yield from 90Sr has been measured as a function of distance (x) from a cerium bromide scintillation detector relative to the 137Cs γ-ray full-energy peak response. The utility of two, count-independent shape factor parameters has been compared: the ratio of the sum of the counts in the 60–800 keV region to the full-energy peak response associated with the 137Cs 662 keV γ-ray line, termed for the benefit of this paper as SF1, and the ratio of counts in the 60–350 keV region to those in the 350–450 keV region, defined in this work as SF2. These parameters probe the presence of isotopes associated with clear full-energy spectral lines and bremsstrahlung, respectively. They have been calibrated as a function of x with sealed sources in the laboratory and both are observed to be asymptotic as x→0. This approach has been tested with a variety of open, laboratory-based liquid samples comprising 90Sr and 137Cs, and in a water-based, contaminated medium in the field associated with a sump at a low-level waste disposal facility on the former fast reactor site at Dounreay, UK. The highest response in both SF1 and SF2 measured in this work is observed in these on-site measurements, consistent with x→0, implying both 90Sr and 137Cs presenting together in solution. This is consistent with prior, laboratory-based analysis of the sump media, and suggests that the β−-emitter contribution (anticipated to comprise 90Sr and 90Y) is very near the probe, consistent with the solution being in contact with it. This is of relevance where the β−/γ composition of contaminated groundwater, in-situ, is not well understood or where migratory changes in activity are suspected.

AI-based detection and identification of low-level nuclear waste: a comparative analysis

Ensuring environmental safety and regulatory compliance at Department of Energy (DOE) sites demands an efficient and reliable detection system for low-level nuclear waste (LLW). Unlike existing methods that rely on human effort, this paper explores the integration of computer vision algorithms to automate the identification of such waste across DOE facilities. We evaluate the effectiveness of multiple algorithms in classifying nuclear waste materials and their adaptability to newly emerging LLW. Our research introduces and implements five state-of-the-art computer vision models, each representing a different approach to the problem. Through rigorous experimentation and validation, we evaluate these algorithms based on performance, speed, and adaptability. The results reveal a noteworthy trade-off between detection performance and adaptability. YOLOv7 shows the best performance and requires the highest effort to detect new LLW. Conversely, OWL-ViT has lower performance than YOLOv7 and requires minimal effort to detect new LLW. The inference speed does not strongly correlate with performance or adaptability. These findings offer valuable insights into the strengths and limitations of current computer vision algorithms for LLW detection. Each developed model provides a specialized solution with distinct advantages and disadvantages, empowering DOE stakeholders to select the algorithm that aligns best with their specific needs.

Multi-Load Topology Optimization Design for the Structural Safety Maintenance of Low- and Intermediate-Level Radioactive Waste Packaging Containers in the Case of a Collision.

This paper presents an optimized design approach using nonlinear dynamic analysis and finite element methods to ensure the structural integrity of square-shaped containers made from ductile cast iron for intermediate- and low-level radioactive waste packaging. Ductile cast iron, with its spherical graphite structure, effectively distributes stress throughout the material, leading to a storage capacity increase of approximately 18%. Considering the critical need for containers that maintain integrity under extreme conditions like earthquakes, the design focuses on mitigating stress concentrations at the corners of square structures. Nonlinear dynamic analyses were conducted in five drop directions: three specified by ASTM-D5276 standards and two additional directions to account for different load patterns. Fractures were observed in four out of the five scenarios. For each direction where fractures occurred, equivalent loads causing similar displacement fields were applied to linear static models, which were then used for multi-load topology optimization. Three optimized models were derived, each increasing the volume by 1.4% to 1.6% compared to the original model, and the design that best met the structural integrity requirements during drop scenarios was selected. To further enhance the optimization process, weights were assigned to different load conditions based on numerical analysis results, balancing the impact of maximum stress, average stress, and plastic deformation energy. The final model, with its increased storage capacity and enhanced structural integrity, offers a practical solution for radioactive waste management, overcoming limitations in previous designs by effectively addressing complex load conditions.

Radionuclide Solid:liquid partitioning in an aged, reducing-grout wasteform recovered from a disposal facility

The Saltstone Disposal Facility on the Savannah River Site in South Carolina disposes of Low-Level Waste in a reducing-grout waste form. Reducing grout is presently being evaluated as a subsurface disposal waste form at several other locations in the United States, as well as in Europe and Asia. The objective of this study was to collect core samples directly from the Saltstone Disposal Facility and measure desorption distribution coefficients (Kd; radionuclide concentration ratio of saltstone:liquid; (Bq/kg)/Bq/L)) and desorption apparent solubility values (ksp; radionuclide aqueous concentration (moles/L)). An important attribute of this study was that these tests were conducted with actual aged, grout waste form materials, not small-volume simulants prepared in a laboratory. The reducing grout is comprised of blast furnace slag, Class F fly ash, ordinary portland cement, and a radioactive salt waste solution generated during nuclear processing. The grout sample used in this study underwent hydrolyzation in the disposal facility for 30 months prior to measuring radionuclide leaching. Leaching experiments were conducted either in an inert (no oxygen) atmosphere to simulate conditions within the saltstone monolith prior to aging (becoming oxidized) or they were exposed to atmosphere conditions to simulate conditions of an aged saltstone. Importantly, these experiments were designed not to be diffusion limited, that is, the saltstone was ground finely and the suspensions were under constant agitation during the equilibration period. Under oxidized conditions, measured Tc Kd values were 10 mL/g, which was appreciably greater than the historical best-estimate value of 0.8 mL/g. This difference is likely the result of a fraction of the Tc remaining in the less soluble Tc(IV) form, even after extensive oxidation during the experiment. Under oxidized and reducing conditions, the measured Ba and Sr (both divalent alkaline earth metals) Kd value were more than an order of magnitude greater than historical best-estimate values of 100 mL/g. The unexpectedly high Ba and Sr Kd values were attributed to these radionuclides having sufficient time to age (form strong bonds) in the sulfur-rich saltstone sample. Apparent ksp values under reducing conditions were 10−9 mol/L Tc and 10−13 mol/L Pu, consistent with values measured with surrogate materials. Measured apparent Ba, Sr, and Th ksp values were significantly greater than historical best-estimates. The implications of the generally greater Kd values and lower ksp values in these measurements is that these cementitious waste forms have greater radionuclide retention than was previously estimated based on laboratory studies using surrogate materials. This work represents the first leaching study performed with an actual aged, reducing-grout sample and as such provides an important comparison to studies conducted with surrogate materials, and provides high pedigree data for other programs around the world evaluating reducing grouts as a wasteform for subsurface nuclear waste disposal.

Selective Adsorption of Sr(II) from Aqueous Solution by Na3FePO4CO3: Experimental and DFT Studies.

The efficient segregation of radioactive nuclides from low-level radioactive liquid waste (LLRW) is paramount for nuclear emergency protocols and waste minimization. Here, we synthesized Na3FePO4CO3 (NFPC) via a one-pot hydrothermal method and applied it for the first time to the selective separation of Sr2+ from simulated LLRW. Static adsorption experimental results indicated that the distribution coefficient Kd remained above 5000 mL·g-1, even when the concentration of interfering ions was more than 40 times that of Sr2+. Furthermore, the removal efficiency of Sr2+ showed no significant change within the pH range of 4 to 9. The adsorption of Sr2+ fitted the pseudo-second-order kinetic model and the Langmuir isotherm model, with an equilibrium time of 36 min and a maximum adsorption capacity of 99.6 mg·g-1. Notably, the adsorption capacity was observed to increment marginally with an elevation in temperature. Characterization analyses and density functional theory (DFT) calculations elucidated the adsorption mechanism, demonstrating that Sr2+ initially engaged in an ion exchange reaction with Na+. Subsequently, Sr2+ coordinated with four oxygen atoms on the NFPC (100) facet, establishing a robust Sr-O bond via orbital hybridization.

Cliff Retreat Rates Associated with a Low-Level Radioactive Waste Disposal Facility in Los Alamos, New Mexico, USA

We present an analysis and interpretation of potential cliff stability at a low-level waste disposal facility at Los Alamos National Laboratory, New Mexico, using cliff morphologic and fracture characteristics coupled with carbon-14 surface exposure dating. Our study is important as it directly bears on the licensing criteria for low-level radioactive waste sites. We find that future characteristic cliff failures will likely not breach disposal pits and shafts over the 1000-year minimum regulatory period. Further, we find, using a multivariate regression model, that slope angle and cliff face aspect are sub-equal in importance to predict regions of high risk of failure when combined with surface exposure ages and assuming that old exposure ages are most indicative of stability (instability) and therefore can aid decision making in final design implementation.

Fabrication of magnetite/GO/potassium copper hexacyanoferrate nanocomposite for removal of radioactive cesium ions

Metal hexacyanoferrates (MHCF) are a class of inorganic adsorbents used for wastewater management due to the presence of interstitial sites for capturing heavy metal ions. In present work, we are reporting the synthesis of magnetic nanocomposite of Fe3O4/graphene oxide/potassium copper hexacyanoferrate via wet chemical and coprecipitation approach. Potassium copper hexacyanoferrate (KCuHCF) and Graphene oxide (GO) both are marvelous adsorbents but their nano-size becomes a major obstacle in their separation process after the adsorption of the radionuclides. Thus, our synthesized nanocomposite Fe3O4/GO/KCuHCF enhances the recovery of KCuHCF even after radioactive Cs+ adsorption with adsorption capacity of 18 mg g−1 coinciding well with the Langmuir adsorption isotherm mechanism. The synthesized adsorbent is characterized thoroughly using UV–Visible spectroscopy, FT-IR, TGA, XPS, Raman spectroscopy, TEM-EDAX and XRD. This synthesized nanocomposite is used for the batch extraction of radioactive Cs+ from low level radioactive waste (LLW). The extraction kinetics followed pseudo-second-order kinetics mechanism.

Handling, Conditioning of Low Level Radioactive Wastes (LLRW), Spent Radiation Sources (SRS), their transport to Temporary Storage Facility in Kosovo and Albania

Handling, Conditioning of Low Level Radioactive Wastes (LLRW), Spent Radiation Sources (SRS), their transport to Temporary Storage Facility in Kosovo and Albania

Effects of Micro- and Nanosilica on the Mechanical and Microstructural Characteristics of Some Special Mortars Made with Recycled Concrete Aggregates.

In this paper, we study the influence of densified microsilica and colloidal nanosilica admixtures on the mechanical strength and the microstructural characteristics of special mortars used for immobilizing radioactive concrete waste. The experimental program focused on the replacement of cement with micro- and/or nanosilica, in different proportions, in the basic composition of a mortar made with recycled aggregates. The technical criteria imposed for such cementitious systems, used for the encapsulation of low-level radioactive waste, imply high fluidity, increased mechanical strength and lack of segregation and of bleeding. We aimed to increase the structural compactness of the mortars by adding micro- and nanosilica, all the while maintaining the technical criteria imposed, to obtain a cement matrix with high durability and increased capacity for immobilizing radionuclides. The samples from all the compositions obtained were analyzed from the point of view of mechanical strength. Also, micro- and nanosilica as well as samples of the optimal mortar compositions were analyzed physically and microstructurally. Experimental data showed that the mortar samples present maximum compressive strength for a content between 6 and 7.5% wt. of microsilica, respectively, for a content of 2.25% wt. nanosilica. The obtained results suggest a synergistic effect of micro- and nanosilica when they are used simultaneously in cementitious compositions. Thus, among the analyzed compositional variants, the mortar composition with 3% wt. microsilica and 2.25% wt. nanosilica showed the best performance, with an increase in compressive strength of 23.5% compared to the control sample (without micro- and nanosilica). Brunauer-Emmett-Teller (BET) analysis and scanning electron microscopy (SEM) images highlighted the decrease in pore diameter and the increase in structural compactness, especially for mortar samples with nanosilica content or a mixture of micro- and nanosilica. This study is useful in the field of recycling radioactive concrete resulting from the decommissioning of nuclear research or nuclear power reactors.

Radiation source terms analysis and classification of radioactive waste for the decommissioning of the first HWRR reactor in China.

Nuclear reactors will face the problem of decommissioning at the end of their operating life due to the high radioactivity of reactor components and environmental safety considerations. The Heavy Water Research Reactor (HWRR) is the first large-scale research reactor to be decommissioned in China. The second phase of HWRR decommissioning involves the main components in the reactor block, so the radiation source terms and the radioactive waste level need to be evaluated before demolition and disposal. Based on the operating history, three-dimensional geometry, materials, and other information of the HWRR, the activity of radionuclides in the main components of HWRR is calculated and analyzed, and the MCNP/ORIGEN coupling scheme is utilized for theoretical analysis. The theoretical results indicate that 14C, 54Mn, 55Fe, 60Co, 63Ni, and 152Eu are the main radioactive nuclides. The total activity of radioactive nuclides was 2.36E + 15Bq at the end of 2007, 4.27E + 13Bq at the end of 2021, and 1.83E + 13Bq at the end of 2025. Furthermore, local sampling and radiometric analyses based on the HPGe gamma-ray spectrometer are also performed to verify the theoretical results, the ratio of theoretical activity values to the measured activity of the experimental sample is within 2.5 times, so the theoretical results are conservative. According to the classification standards for radioactive waste, the inner shell, outer shell, cooling water tank, sand layer, and heavy concrete shielding layer are all low-level waste. These results and conclusions can serve as a reference for the second phase decommissioning of the HWRR and the subsequent disposal of radioactive waste.