Hasil untuk "Nuclear engineering. Atomic power"

Yang Yang, Haiqing Deng, Liang Zhu et al.

Precise calibration of Motional Stark Effect (MSE) diagnostic sightline is fundamental for reliable plasma current density measurements in EAST tokamak. This paper presents the development and experimental validation of a vision-based robotic system for automatic MSE calibration. The proposed system integrates a 6 degree-of-freedom industrial manipulator with computer vision techniques to achieve three key advancements: (1) automated measurement of sightline pose, (2) polarized light generation and alignment, and (3) synchronized acquisition of optical transmission system responses. The calibration robot system was validated by workspace simulation, test bench in laboratory and practical deployment for MSE calibration within the EAST vacuum vessel during the 2024 and 2025 experimental campaigns. Compared to traditional calibration methods (based on 2021 and 2023 campaign data), the proposed robotic calibration system demonstrates significantly improved performance, by reducing the average sightline pose measurement repeatability deviation from 12.1 mm to 2.0 mm.

Evgeniy Adamov, B. Gabaraev, N. Gorin et al.

Xu Wu, L. Moloko, Pavel M. Bokov et al.

Machine learning (ML) has been leveraged to tackle a diverse range of tasks in almost all branches of nuclear engineering. Many of the successes in ML applications can be attributed to the recent performance breakthroughs in deep learning, the growing availability of computational power, data, and easy-to-use ML libraries. However, these empirical successes have often outpaced our formal understanding of the ML algorithms. An important but under-rated area is uncertainty quantification (UQ) of ML. ML-based models are subject to approximation uncertainty when they are used to make predictions, due to sources including but not limited to, data noise, data coverage, extrapolation, imperfect model architecture and the stochastic training process. The goal of this paper is to clearly explain and illustrate the importance of UQ of ML. We will elucidate the differences in the basic concepts of UQ of physics-based models and data-driven ML models. Various sources of uncertainties in physical modeling and data-driven modeling will be discussed, demonstrated, and compared. We will also present and demonstrate a few techniques to quantify the ML prediction uncertainties. Finally, we will discuss the need for building a verification, validation and UQ framework to establish ML credibility.

Trong-Ha Nguyen, D. Nguyen

Base isolators constitute solutions for improving the seismic performance of civil and nuclear engineering structures. This paper evaluates the seismic fragility of based-isolated nuclear power plant structures using the proposed fragility curves. A finite element model of the structures is developed deploying SAP2000, a structural analysis program. For constructing fragility curves, a set of ground motions is employed to perform nonlinear time-history analyses associated with Incremental Dynamic Analyses (IDA). Three Damage States (DS) are defined based on the shear deformation of base isolators. Finally, the maximum likelihood estimation technique generates a set of fragility curves for DS. Additionally, a comparison of fragility curves between IDA and Cloud Analysis (CA) is presented.

Fred D. Lang

Abstract This work proposes a paradigm shift in nuclear safety. Its NCV Method (neutronics/calorimetrics/verification procedures) integrates nuclear power’s motive force—neutron flux—within Second Law exergy analysis, coupled with corrected conservation of energy flows, both descriptive of the entire system. These descriptions with two satellite equations result in a verifiable understanding of the nuclear engine: neutron flux [or neutronic terms f (Φ)], useful power produced, and system heat rejection, all coupled to reactor vessel coolant mass flow. Key to NCV is its assumption that all nuclear phenomena are inertial processes, devoid of terrestrial reference. This approach demands reinterpretation of Einstein’s ΔE = c2Δm by describing his ΔE as an exergetic potential, an ultimate Free Exergy. For fission, Free Exergy consists of both recoverable and irreversible portions given a total MeV release. In transference to the coolant, the recoverable release produces an exergetic increase ($$\rm \dot m$$m˙Δg) in the fluid; an explicitly calculated Inertial Conversion Factor produces a computed Core Thermal Power ($$\rm \dot m$$m˙Δh) and a nuclear TRef. This paper asserts that traditional nuclear engineering has lacked direct linkage between neutron flux and system fluid thermodynamics. With NCV, nuclear power’s motive force is explicitly related to extensive properties, thus allowing reconciliation of principal system parameters of the nuclear engine (fission source, power out, heat rejection, and main system flow). Principal verification is accomplished by comparing the computed useful power to that which is directly measured. The NCV method has the potential to reduce uncertainty in computed Core Thermal Power from its commonly accepted ±2% by an order of magnitude. Its ability to improve plant safety becomes intrinsic, for example:  ● tracking changes in verifiable flux versus reactor vessel coolant flow;  ● tracking changes in the axial position where saturation may be approached, and the position of the average coolant temperature;  ● monitoring the instantaneously computed flux versus reactor vessel pump currents;  ● detecting changes in the important ΔλGEN and ΔλEQ40 verification parameters, which compare the computed shaft power delivered to the generator, to the measured;  ● surveilling component irreversible losses using Fission Consumption Indices; etc.

Domen Kotnik, Julijan Peric, Domen Govekar et al.

Water as a primary coolant will play an important role in the performance of fusion reactors, as it causes an ionising radiation field throughout the facility after its irradiation and activation and requires improved shielding for instruments and personnel. To support ITER, the KATANA irradiation facility, which utilises a closed-water activation loop, was built, licenced and commissioned in early 2024 at the TRIGA research reactor at the Jožef Stefan Institute in Slovenia.A comprehensive overview of the KATANA irradiation facility, which serves as a well-defined and stable high-energy γ and neutron source, is given, including the final design, experimental set-up, detector systems and overall timeline. The physical characteristics of KATANA were assessed by analyses of neutron transport simulations leading to activity calculations in connection with a conventional analytical approach, without using CFD calculations. The KATANA facility demonstrated the desired operational characteristics in terms of high and stable water flow rates, leading to high activity values of the observed isotopes 16N, 17N and 19O, which is essential for minimising experimental uncertainties.The preliminary calculations provided in-depth operational knowledge and important data for the KATANA facility, forming a basis for more advanced future experiments.

Jian-song Zhang, Hua-ping Mei, Yong-ju Sun et al.

As small modular reactors continue to advance, heat pipe reactors (HPRs) are being developed by multiple organizations. Designing heat pipes with optimal heat transfer capability is crucial for the effective performance of HPRs. Based on NSGA-III genetic algorithm, this paper presented the design method of a high-temperature heat pipe featuring a composite wick. The study examined the influence of heat pipe structure parameters on performance parameters and heat transfer limits. Based on the analysis, we built multi-objective modeling of heat pipe and obtained the optimal structural parameters by NSGA-III. Verification confirmed that the heat pipe met the Mach number and effective capillary radius requirements. This study provided valuable insights for improving heat pipe manufacturing techniques and reactor miniaturization.

Yikai Wang, Chenxiao Tang, Liming Tan et al.

The nuclear power plant's rotating asynchronous machine (RAM) is the sole power supply system of the control rod drive mechanism (CRDM). The current RAM's protection system consists of traditional discrete relays, which have low reliability and flexibility. The popularization of digital protection devices and new principles has been started in the engineering field. Based on the feature of the Zig-Zag star winding connection of the generator stator winding in RAM, this paper analyzes the after-ground fault third-harmonic voltage characteristics. It is found that when the fault resistance exceeds 5 kΩ, the application of the traditional third harmonic voltage criteria is difficult because of the low sensitivity. Thus, this paper proposes a novel protection method based on the variation in the third-harmonic voltage, effectively improving protection reliability by introducing restraining quantity. Simulation and dynamic model experiments show that, compared to the existing third-harmonic voltage criteria, the proposed method exhibits higher sensitivity and has an against fault resistance ability of at least 10 kΩ.

Geun-Ho Kim, Ilgook Kim, Kwang Pyo Kim

To obtain site-specific values of the Derived Concentration Guideline Levels (DCGLs) for decommissioning of KRR-1&2, the soil density and distribution coefficient values for Cs-137, a major contaminant radionuclide, were determined. The soil density was evaluated according to the test method established by the Korean Agency for Technology and Standards of the Ministry of Trade, Industry, and Energy (KATS). The distribution coefficient was evaluated using a batch test. The validity of using the evaluated soil density and distribution coefficient as site-specific values was assessed through radiation dose assessment reflecting these values. Average soil density value obtained was 1.738 g/cm3, which was within the typical range of normal soil density, 1.0–1.8 g/cm3. The average distribution coefficient value was 7,754 mL/g. Applying the maximum, average, and minimum values of the evaluated soil density and distribution coefficient showed similar radiation dose results, thus suggesting that it is reasonable to use the average values of each parameter as site-specific values. Findings of this study can help determine DCGLs that reflect the characteristics of the research reactor site.

Yanfei Zhao, Yan Zhang, Yude Huang et al.

Ionizing radiation can cause damage to the structure and function of submandibular glands through oxidative stress and apoptosis. The polysaccharide of Dicliptera chinensis is a kind of natural antioxidant, which has the radiation protection effect and can prevent the damage caused by ionizing radiation. The aim of this study is to investigate the protective effect of Dicliptera chinensis polysaccharide on radiation-induced submandibular gland injury and its mechanism. Sprague-Dawley rats were irradiated with 18Gy X-rays and treated with Dicliptera chinensis polysaccharide (200 mg/kg) or amifostine (250 mg/kg). In this study, the changes of salivary gland secretion function were observed by measuring saliva flow rate, the histamathological changes of submandibular gland were observed by hematoxylin-eosin staining and immunohistochemistry, the expression of oxidative stress-related indicators (ROS, T-AOC) and their markers (NOX4, SOD2) in submandibular gland tissues were detected by ELISA, TUNEL and WB experiments, and the expression of apoptosis-related proteins (p-p53, Bax, Bcl2, caspase 9, caspase 3) was detected by western blotting assay. The results showed that Dicliptera chinensis polysaccharide protected salivary gland secretory function by alleviating the damage of AQP-5 protein on submandibular gland epithelial cell membrane, reduced the oxidative stress injury of submandibular gland by regulating the expression of NOX4 and SOD2 proteins. Furthermore, Dicliptera chinensis polysaccharide reduced apoptosis by regulating the p53-bax/bcl2 pathway. In summary, Dicliptera chinensis polysaccharide could be used as a radiation protection agent to prevent radiation damage of submandibular glands in the future.

SHAN Yuchang, LIU Xiaojing, HE Hui

BackgroundTetragonal BaTiO3 exhibits temperature-dependent surface wettability as a pyroelectric material, hence is expected to be exploited to improve boiling and heat transfer efficiency on the surface of nuclear-reactor heat exchange components.PurposeThis study aims to prepare BaTiO3 pyroelectric thin films and explore their properties.MethodsFirstly, TiO2 nanotubes were prepared by anodic oxidation, and a controllable preparation of BaTiO3 nanotube array films was achieved using hydrothermal synthesis. Then, the X-ray Diffraction (XRD) and Scanning Electron Microscope (SEM) were employed to observe characteristics and analyze the growth mechanism of TiO2 nanotube and BaTiO3 nanotube array film. Finally, the surface morphology and phase structure changes of the nanotubes were investigated by adjusting the voltage, NH4F concentration, and oxidation time.ResultsThe results show that the size of the generated oxygen bubbles increase with the increase of electron current caused by high voltage, and the diameter of the nanotubes increases with the oxidation voltage. The tube diameter distribution ranges within 60~140 nm, and the tube wall thickness is 10 nm. Increasing the concentration of NH4F and oxidation time are beneficial for the formation of TiO2 nanotubes. Polishing the titanium sheet can considerably improve the flatness of the nanotube array generated by oxidation. By extending the hydrothermal time and increasing the high-temperature annealing treatment, the cubic phase of BaTiO3 is successfully converted into a tetragonal phase with pyroelectric effects. Compared with the sample prepared over longer hydrothermal time, the annealed sample exhibits better pyroelectric properties.ConclusionsThe results of this study provide a valuable reference for further analyzing the growth mechanism of anodized TiO2 nanotubes and exploiting the spontaneous polarization intensity change of pyroelectric materials to change the surface wettability and improve the boiling heat transfer.

Soohyeok Lee, Hyoungtaek Kim, Hwijoon Jung et al.

This paper explores the implementation of machine learning-based algorithms for TL dose assessment. It focuses on the radiation field classification, performance quotient evaluation, and shallow and deep dose equivalent assessment of ANN and LGBM, in comparison to the traditional method of DT. We evaluate these algorithms based on the element response data measured by TLD. A data set was built for training, and the base element responses of test categories were amplified, and normalized to 1 mSv Cs-137 within the range of ±3 %. Both algorithms consist of five subset models for classifying radiation fields and identifying ratios of mixed fields. The LGBM showed the best accuracy in classifying considered radiation fields and the lowest performance quotients. By comparing the tolerance levels of deep dose and shallow dose equivalents among the three algorithms, the LGBM yields the smallest difference between the predicted and true dose equivalents. This smaller difference implies the LGBM offers the least bias and standard deviation in the expected value, giving higher accuracy and precision in dose assessment over the traditional DT method. The findings from this study further contribute to the adoption of ML-based algorithms for TL dose assessment, underscoring its importance in the field.

V. A. Vinniychuk, N. N. Podrezov, Yu.V. Doronin

Welding equipment is everywhere equipped with modern digital welding arc power systems based on microprocessor and other logical element bases. At the same time, there is a tendency to minimize information not only for Chinese, but also for equipment of well-known brands, so it is important to quickly monitor the accuracy of the accompanying documentation in order to avoid deterioration in the quality of products and waste of time re-equipment. Modern recorders of welding processes allow several times to reduce the time for checking the declared characteristics of all power sources certified for use in the nuclear power industry. Among the characteristics and service functions analyzed by the recorder, it should be noted static current-voltage characteristics, oscillograms of «hot start», «afterburner» and «anti-sticking». The paper presents the data of selective tests for compliance with the declared passport data of inverter power supplies INEM-200T (MMA process), Artsen CM-500 (MAG process) and ION 48-900 (SAW process). Based on the results of the research, conclusions were drawn about the benefits of using the proposed model of the welding process recorder at enterprises of heavy nuclear engineering.

Yuyang Ren, Hao Yao, Guofang Liu et al.

In this paper, a text code recognition and positioning system for engineering drawings of nuclear power equipment is designed to promote the informatization and automation of paper work created in nuclear power projects including construction procedure and maintenance procedure. Its core function is to analyze engineering drawings, and identify the user-searched code contents in them, and then extract the codes and related information, so that the staff can more quickly retrieve the required codes, obtain the drawing files containing the codes, and locate the codes on the engineering drawings precisely. The code recognition system proposed in this paper is based on the Teigha toolkit, ezdxf open source library, CRNN and PaddleOCR open source library. For engineering drawings in picture format, VGG16 is used for feature extraction of code regions. CRNN based on VGG16 is used to divide the entire coding area, and PaddleOCR is used to identify the coding information. The experiment results show that the system can be effectively used in engineering practice of nuclear power project with high usability, which can greatly improve the work efficiency of staff.

Hao Wan, Songfeng Weng, Hua Du et al.

The transportation of heavy equipment in nuclear engineering has always been the focus of engineers, especially those transfer devices with the characteristics of small geometric size and heavy load. According to this kind of compact heavy-load transfer device and its engineering tasks, the core problems caused by excessive vertical and horizontal forces in the design process were analyzed. By introducing the theory of inventive problem solving (TRIZ) design method, these problems were creatively solved by the contradiction theory and substance-field model in TRIZ, and an innovative design scheme of the compact heavy load-independent transfer device was obtained. Through the analysis of the design scheme and the stability and rapidity of its hydraulic system, some key parameters were determined. The power of the transfer device was all from the hydraulic system, and it can carry up to 300 t weight of reactor equipment, while its geometric size was only 1600 × 400 × 500 mm. It was of great significance to improve the efficiency of the nuclear engineering system.

S. Dudarev

The 1992 United Nations Rio de Janeiro declaration (United Nations, 1992) states that “Human beings are at the centre of concerns for sustainable development. They are entitled to a healthy and productive life in harmony with nature.” This brief affirmation highlights the implications from the impact of growing human population on the environment (Cartledge, 1995; MacKay, 2009), manifested in the notion of climate emergency (Ripple et al., 2020). Nuclear power offers progressive options to mitigate global warming and other effects of climate change, with the International Energy Agency (IEA) suggesting that the nuclear energy generation currently eliminates between 1.3 and 2.6 giga-tonnes of CO2 emissions from the power sector each year, depending on whether it is assumed that it replaces gasor coal-fired power plants. The IEA’s 2015Technology Roadmap (IEA, 2015) report noted that tomeet the Paris Agreement target of global temperature not rising by more than 2°C, the world nuclear power generation capacity needs to increase to 930 GW in 2050. For comparison, the Smart Energy Europe analysis projected the European nuclear power generation capacity in 2050 to the level of 105 GW (Connolly et al., 2016), against the current levels of 61.3 GW in France, 9 GW in the United Kingdom, raising to 24 GW by 2050 (UK Government, 2022), and 4.3 GW in Germany, down from the 2021 level of 7.4 GW. Comparing these figures and the data for other geographical areas with the projected world total nuclear generation capacity, we observe that many new power plants are expected to be constructed in the countries where nuclear power generation technology and engineering have so far been largely unknown. This expected expansion will involve a broad range of engineering and technological challenges spanning the manufacturing of reactor components, the fabrication and extraction of fuel, the development of efficient coolant and heat transfer technology, the reactor assembly schedules, the establishment of supporting hot cell and waste processing facilities, and the technology for scheduled and unscheduled remote maintenance and operation, enabling the reactor systems to reliably function over long periods of time. All the presently operating commercial nuclear reactors use fissile nuclear fuel, containing isotopes of uranium and other actinide elements. On the other hand, fusion power generation, an area of active development and innovation worldwide, aims to use light fusable chemical elements, for example the deuterium and tritium isotopes of hydrogen (Pearson and Takeda, 2020; Prager and Najmabadi, 2020). Fusion technology presents a range of scientific and engineering challenges that need to be addressed to enable the construction of a fusion power plant (Chapman and Walkden, 2020). These include the development of a reliable and safe tritium and deuterium extraction and handling technology, the integration of structural and functional materials in a power plant design, and the extensive use of remote handling and robotics in the maintenance of a power plant. But first and foremost, it is the development of robust means for controlling the high temperature plasma (Kodama et al., 2001; Hender et al., 2007), either in a magnetic confinement device or in a pulsed, for example a laser-driven, fusion system that presents an outstanding challenge to the fusion power plant engineering. The fundamental considerations involved in the assessment of nuclear power are its economic competitiveness against the power sources using coal and gas, or the renewable sources like solar and wind power (Alonso et al., 2016), and its environmental impact (Pigford, 1974; Danish et al., 2022). Edited and reviewed by: Edgar C. Buck, Pacific Northwest National Laboratory (DOE), United States

SUN Qi;WANG Zhaohui;ZHANG Qiwei;HUANG Hanxiong;REN Jie;RUAN Xichao;LIU Shilong;BAO Jie;LUAN Guangyuan;DING Yanyan;CHEN Xiongjun;NIE Yangbo;LIU Chao;ZHAO Qi;WANG Jincheng;HE Guozhu;DU Shubin

Bismuth is widely used in advanced nuclear reactors such as lead-bismuth eutectic (LBE) fast reactors, space reactors and so on. Its neutron nuclear data, especially the inelastic scattering cross sections have significant impact on safety and economics of these nuclear facilities. A facility for measurement of neutron reaction cross sections using prompt γ ray method based on the HI13 tandem accelerator was established at China Institute of Atomic Energy (CIAE). Neutrons were produced using a D2 gas target bombarded with deuterons at 65, 80 and 95 MeV. The neutron target was positioned in the center of a shielding in size of 2 m×2 m×2 m composed of concrete, iron, lead and borated polyethylene. Neutrons were extracted at 0° with respect to the deuteron beam through a collimator made of copper, iron, polyethylene and lead. Neutron inelastic scattering cross sections at 90, 105 and 120 MeV were measured experimentally for 209Bi with a 209Bi sample in size of 50 mm×4 mm. A natural titanium sample in size of 50 mm×1 mm was used as the reference sample. By measuring production cross sections of the 983.5 keV γ rays produced when neutrons scattered inelastically off 48Ti nuclei, a normalization factor can be determined. 4 Clover detectors were used to measure produced γ rays. They were placed at 30°, 70°, 110° and 150° with respect to the deuteron beam. Lead shielding in thickness of approximately 2 cm was used for detector head to shield scattered γ rays. Timeofflight method was used to determine neutron energy, thus neutrons produced by DD breakup could be discriminated with neutrons produced by DD fusion. The energy and absolute efficiency were calibrated using 152Eu, 60Co, 22Na and 133Ba standard γ sources with known radioactive activities. A 508 cm×508 cm liquid scintillation detector was positioned at the end of neutron beam line to monitor the neutron fluxes passing through the samples. Two specific measurement angles at 110° and 150° were selected to obtain angleintegrated γ production cross sections by performing a simple linear summation of the partial cross sections measured at these angles. Talys 195 code was used to calculate ratios of γ production cross sections and total inelastic scattering cross sections. The experiment results were compared with other experiment data, data retrieved from ENDF/BⅧ0, JEFF33, JENDL40, ROSFOND2010 and CENDL31 evaluated nuclear data libraries as well as calculation by Talys 195 code with the default parameters. The results show that the tendency of these 3 measured energy points is similar to these results. For cross sections measured at 90 and 105 MeV, the results are closer to data calculated with Talys 195 code. For cross section measured at 120 MeV, it fits the ROSFOND2010 evaluated data better.

Mizuki Sakamoto, Yaowei Yu, Naoko Ashikawa et al.

Deuterium recycling properties have been investigated in the EAST superconducting tokamak by evaluating the density decay time after the fueling termination. The density decay time of the latter discharge of the experiment of a day is ∼2.9 s, which is more than three times longer than that of the beginning discharge (∼0.88 s), indicating that a repeat of discharges diminishes the wall conditioning effect by lithium coating. However, the difference in deuterium recycling among three magnetic configurations (lower single null, double null, and upper single null) is unclear. The density decay times are 3.3 s and 4.1 s when the strike point (SP) positions are located on the horizontal and vertical plates of the lower outer divertor, showing that the particle exhaust capability becomes stronger as the SP position approaches the pumping slot. Furthermore, the lower hybrid current drive (LHCD) power enhances deuterium recycling since the density decay times of high (∼2.7 MW) and low (∼0.8 MW) power LHCD discharges are 4.3 s and 1.2 s, respectively. The density decay time in the LHCD discharge linearly increases with an increase in a period of exponential fitting, implying that the recycling coefficient increases during the stop of fueling. However, the density decay time in the ohmic discharge is independent of the fitting time.

Wei Zheng, Rong Yan, Rui Ding et al.

Material erosion and deposition are one of the important topics related to plasma-wall interaction in tokamaks. Detailed analysis of the deposits on the surface of the graphite tile from the divertor dome region has been made after the 2019 experimental campaign. It is indicated that the deposits exhibited porous globular in morphology which were mainly stacked by columnar grains with size of 2–5 μm and small crystal particles with several hundreds of nanometers in size. The thickness of the deposits varied from several micrometers to about 120 μm. As the SiC coating on the graphite tile has a rough surface and exhibited valleys and hills, the deposits on the valleys were typically thicker than that on the hills. The deposits consisted of Li, C, O, W, Mo, Fe, Cu, Ni, et. mainly from the wall conditioning materials and plasma-facing materials. It was noted that Li was in the form of Li2CO3. And Mo and W were in MoO3 and WO3 respectively. The total content of Li2CO3 was found to be higher than 90 wt% with a corresponding Li density in deposits of about 7.22 μg/mm−2 in average due to the routine Li wall conditioning. The depth profile of Li exhibited differently from the other elements. And W has a higher density than Mo, which could be due to the high W erosion from the upper divertor.

Yi-Sin Chou, Bhupendra Singh, Yong-Song Chen et al.

Perlite is one of the major constituents of the radioactive thermal insulation waste (RTIW) originating from nuclear power plants and, for proper waste management, a significant reduction in its volume is required prior to disposal. In this work, the volume reduction of perlite is studied by high-temperature treatment method with using K2CO3 as a flux. The perlite is ground with 0–30 wt% K2CO3, and differential thermal analysis/thermogravimetric analysis is used to monitor the glass transition temperature (Tg) and weight loss. The Tg varied between ~772.2 and 837.1 °C with the minima at ~643.5 °C with the addition of ~10 wt% K2CO3. It is observed that compared to the pure perlite the volume reduction ratio (VRR) increases with the addition of K2CO3. The VRR of 11.20 is observed with 5 wt% K2CO3 at 700 °C, as compared to VRR of 5.56 without K2CO3 at 700 °C. The X-ray photoelectron spectroscopy and scanning electron microscopy are used to characterize perlite samples heat-treated without/with 5 wt% K2CO3 at 700 °C. Moreover, the atomic absorption spectroscopy indicates that the proposed heat-treatment procedure is able to completely retain the radionuclides present in the perlite RTIW.