Gas-cooled Reactor Design Research Articles

Desain High Temperature Gas-Cooled Reactor (HTGR) Model Geometri Heksagonal Dua Dimensi dengan Bahan Bakar Thorium Hasil Daur Ulang

The Research about the design of high temperature helium gas-cooled reactor (HTGR) terraces with thorium fuel recycled using the SRAC program has been completed. This research includes the percentage of fuel enrichment, reactor core size, reactor core configuration, criticality, and the distribution of the power density. The calculation of reactor core is done in two dimensions \sfrac{1}{6} hexagonal terrace section with a triangular mesh. The fuel is used, i.e. thorium with a burn-up of 20 GWd/t and 30 GWd/t, and helium gas as a cooler. The results obtained in this study show that the ideal HTGR reactor core design with reactor core size and configuration are (x) 22 cm at point (y) = 2035,05 cm and at (y) 11 cm at point (x) = 2035,05 cm, then enrichment in fuel 8%. The result of maximum power density is 550.3685 Watt/cm3 where the position at (x) = 22 cm and axis (y) = 11 with the effective multiplication factor value keff of 1,0000002.

The challenges of gas-cooled reactor technology for space propulsion and the development of the JANUS space reactor concept

There is a strong motivation to develop high-power output nuclear fission reactors (around 1 MWe) for space applications, such as high-payload missions and long-duration missions beyond Mars, where the reduced solar flux makes using alternative energy sources challenging. Many published gas-cooled reactor designs for space applications deliver outputs in the 100 kWe regime, with typical power densities of around 4 MW/m3. Here we present a gas-cooled reactor design – referred to as JANUS – employing He–Xe coolant and operating on a direct Brayton cycle, that can achieve a high-power output (around 0.8 MWe) and higher power density (24 MW/m3) than previously published designs. The core employs a coated particle fuel form with uranium nitride kernels in a graphite matrix and with a high packing fraction (45%).Starting from the CEA's published OPUS space reactor design, a variety of approaches were considered for achieving high-power densities in a gas-cooled reactor, including changing the core aspect ratio. Our JANUS design uses a decreased coolant channel diameter and fuel element pitch, whilst increasing the number of fuel elements and employing radial enrichment zoning. The design achieves a significant increase in power density whilst obeying core-life limits of 2000 Effective Full Power Days and a peak fuel operating temperature of 1900 K. The core has been modelled neutronically with the ERANOS and SERPENT codes, with a simple yet robust thermal dimensioning tool developed by us to study peak fuel temperatures, to ensure the fuel operates within its design limitations and to aid optimisation.

POSCA: A computer code for fission product plateout and circulating coolant activities within the primary circuit of a high temperature gas-cooled reactor

Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy.

A new approach for crack detection and sizing in nuclear reactor cores

Remote Visual Inspection (RVI) of reactors in nuclear power plants allows station operators to assess the health and condition of their plant. In the UK, most nuclear stations are of the Advanced Gas-cooled Reactor (AGR) design. During planned periodic outages, a representative portion of each AGR core is inspected using specialist tools equipped with various sensors including a video camera for RVI. If cracks are observed in the core during data capture, a stitched image of the region needs to be created so that the crack can be analysed and sentenced (classifying the crack morphology, location, orientation and size) before the station is returned to service, provided return to service is justified. Currently, the crack analysis and sizing activities are conducted manually by expert analysts in a laborious process. In this paper, we present a new image processing approach capable of automating aspects of the crack analysis process. Specifically, we describe a set of techniques for quickly and accurately detecting the presence of cracks in AGR fuel channel inspection images. We also present a method for detecting circular channel features known as trepanned holes whose dimensions are known and can thus be used for scaling. The results of applying the proposed techniques are evaluated on image data from real AGR fuel channels and are shown to produce comparable results to those obtained manually. The advantage of the proposed approach is that it is fast, robust and more repeatable than the existing manual approach.

Self-shielding effect of double heterogeneity for plutonium burner HTGR design

The investigation on self-shielding effect of double heterogeneity for plutonium burner High Temperature Gas-cooled Reactor (HTGR) design has been performed. Plutonium burner HTGR designed in the previous study by using the advantage of double heterogeneity to control excess reactivity. In the present study, the mechanism of the self-shielding effect is elucidated by the analysis of burn-up calculation and reactivity decomposition based on exact perturbation theory. As a result, it is revealed that the characteristics of burn-up reactivity are determined by resonance cross section peak at 1 eV of Pu-240 due to the surface term of background cross section, this is, the characteristics of neutron leakage from fuel lump and collision to a moderator. Moreover, significant spectrum shift is caused during the burn-up period, and it enhances reactivity worth of Pu-239 and Pu-240 in EOL.

Machine Learning Model for Event-Based Prognostics in Gas Circulator Condition Monitoring

Gas circulator (GC) units are an important rotating asset used in the advanced gas-cooled reactor design, facilitating the flow of CO $_2$ gas through the reactor core. The ongoing maintenance and examination of these machines are important for operators in order to maintain safe and economic generation. GCs experience a dynamic duty cycle with periods of nonsteady state behavior at regular refueling intervals, posing a unique analysis problem for reliability engineers. In line with the increased data volumes and sophistication of available technologies, the investigation of predictive and prognostic measurements has become a central interest in rotating asset condition monitoring. However, many of the state-of-the-art approaches finding success deal with the extrapolation of stationary time series feeds, with little to no consideration of more complex but expected events in the data. In this paper, we demonstrate a novel modeling approach for examining refueling behaviors in GCs, with a focus on estimating their health state from vibration data. A machine learning model was constructed using the operational history of a unit experiencing an eventual inspection-based failure. This new approach to examining GC condition is shown to correspond well with explicit remaining useful life measurements of the case study, improving on the existing rudimentary extrapolation methods often employed in rotating machinery health monitoring.

Bypass Flow Resistance in Prismatic Gas-Cooled Nuclear Reactors

Available computational fluid dynamics (CFD) predictions of pressure distributions in the vertical bypass flow between blocks in a prismatic gas-cooled reactor (GCR) have been analyzed to deduce apparent friction factors and loss coefficients for nuclear engineering systems and network codes. Calculations were performed for vertical gap spacings “s” of 2, 6, and 10 mm — representing 1, 3, and 5 mm in a GCR design, horizontal gaps between the blocks of 2 mm and two flow rates, giving a range of vertical gap Reynolds numbers ReDh of about 40–5300. The present focus is on the examination of the flow in the vertical gaps. Horizontal gaps are treated in CFD calculations but their flows are not examined. Laminar predictions of the fully developed friction factor ffd were about 3–10% lower than the classical infinitely wide channel. In the entry region, the local apparent friction factor was slightly higher than the classic idealized case, but the hydraulic entry length Lhy was approximately the same. The per cent reduction in flow resistance was greater than the per cent increase in flow area at the vertical corners of the blocks. The standard k–ϵ turbulence model was employed for flows expected to be turbulent. Its predictions of ffd and flow resistance were significantly higher than direct numerical simulations (DNS) for the classic case; the value of Lhy was about 30 gap spacings. Initial quantitative information for entry coefficients and loss coefficients for the expansion–contraction junctions between blocks is also presented. The present study demonstrates how CFD predictions can be employed to provide integral quantities needed in systems and network codes.

Extension of double heterogeneity treatment method for coated TRISO fuel particles

This paper extends the methodology for treating the resonance self-shielding effect of doubly heterogeneous (DH) tristructural-isotropic fuel particles to take the coating layers of fuel grains into account. Due to the two-region assumption of fuel kernel and graphite matrix in the original derivation of Williams’ method, which was developed to replace the time consuming SCALE continuous energy spectrum calculations with more efficient equivalence theory and intermediate resonance approximation, the method can be applied only to the fuel kernels without coating layers. In order to overcome this limitation, Ji’s analytical expression of the Dancoff factor for multi-layer spheres has been incorporated into the DH method and the applicability has been successfully extended to the kernels with coating layers. It is demonstrated that the new method can predict the DH effects within 200 pcm order of errors against the reference Monte Carlo calculations for wide range of block type gas cooled reactor designs with coating layers.

Study on air ingress of the 200 MWe pebble-bed modular high temperature gas-cooled reactor

Air ingress, one of the beyond design basis accidents (BDBA) for high temperature gas-cooled reactors (HTR), receives high attention during the design of the 200MWe pebble-bed modular high temperature gas-cooled reactor (HTR-PM), because it may result in severe consequence including the corrosion of the fuel elements and graphite reflector. A hypothetical air ingress accident, caused by the double-ended guillotine break (DEGB) of the horizontal coaxial hot-gas duct, is studied in this paper.After the rupture of the hot-gas duct, the following accident developing progress can be divided into three stages: (1) depressurization, (2) slow air ingress via diffusion and local convection, (3) onset of the overall, stable natural circulation and the consequent continuous air ingress. Based on the design of the HTR-PM, the time scale of diffusion and convection in the second stage is studied to evaluate the onset time of the stable natural circulation. Furthermore, the TINTE model of the HTR-PM is then established to simulate the hot-gas duct DEGB accident, including the flow rate of natural circulation, the corrosion of the fuel elements and the graphite reflector, the feasibility of the mitigation measures. The results show that due to the high pebble bed core with large flow resistance, as well as the complicated flow structure of the HTR-PM, it would take long time to obtain stable natural circulation. The results also indicate that, the bottom reflector graphite could consume most of the oxygen before it reaches the core and the injection of the nitrogen can effectively mitigate the corrosion of the graphite.

The IAEA coordinated research programme on HTGR uncertainty analysis: Phase I status and Ex. I-1 prismatic reference results

The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied, in contrast to the historical approach where sensitivity analysis were performed and uncertainties then determined by a simplified statistical combination of a few important input parameters. New methodologies are currently under development in the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity. High Temperature Gas-cooled Reactor (HTGR) designs require specific treatment of the double heterogeneous fuel design and large graphite quantities at high temperatures. The IAEA has therefore launched a Coordinated Research Project (CRP) on HTGR Uncertainty Analysis in Modelling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain.Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250MW modular pebble bed design similar to the Chinese HTR-PM. Work has started on the first phase and the current CRP status is reported in the paper. A comparison of the Serpent and SCALE/KENO-VI reference Monte Carlo results for Ex. I-1 of the MHTGR-350 design is also included. It was observed that the SCALE/KENO-VI Continuous Energy (CE) k∞ values were 395pcm (Ex. I-1a) to 803pcm (Ex. I-1b) higher than the respective Serpent lattice calculations, and that within the set of the SCALE results, the KENO-VI 238 Multi-Group (MG) k∞ values were up to 800pcm lower than the KENO-VI CE values. The use of the latest ENDF-B-VII.1 cross section library in Serpent lead to ∼180pcm lower k∞ values compared to the older ENDF-B-VII.0 dataset, caused by the modified graphite neutron capture cross section. The fourth beta release of SCALE 6.2 likewise produced lower CE k∞ values when compared to SCALE 6.1, and the improved performance of the new 252-group library available in SCALE 6.2 is especially noteworthy. A SCALE/TSUNAMI uncertainty analysis of the Hot Full Power variant for Ex. I-1a furthermore concluded that the 238U(n,γ) (capture) and 235U(v¯) cross-section covariance matrices contributed the most to the total k∞ uncertainty of 0.58%.

ICONE23-1589 HEAT TRANSPORT ANALYSIS IN A DISTRICT HEATING AND SNOW MELTING SYSTEM IN SAPPORO AND ISHIKARI, HOKKAIDO APPLYING WASTE HEAT FROM GTHTR300

A district heating and snow melting system utilizing waste heat from Gas Turbine High temperature Gas Reactor of 300 MWe (GTHTR300), a heat-electricity cogeneration design of high temperature gas-cooled reactor, was analyzed. Application areas are set in Sapporo and Ishikari, the heavy snowfall cities in Northern Japan. The heat transport analyses are carried out by modeling the components in the system; pipelines of the secondary water loops between GTHTR300s and heat demand district and heat exchangers to transport the heat from the secondary water loops to the tertiary loops in the district. Double pipe for the secondary loops are advantageous for less heat loss and smaller excavation area. On the other hand, these pipes has disadvantage of more electricity consumption for pumping. Most of the heat demand in the month of maximum requirement can be supplied by 2 GTHTR300s and delivered by 9 secondary loops and around 5000 heat exchangers. Closer location of GTHTR300 site to the heat demand district is largely advantageous economically. Less decrease of the distance from 40 km to 20 km made the heat loss half and cost of the heat transfer system 22% smaller.

Effect of bed configuration on pebble flow uniformity and stagnation in the pebble bed reactor

Pebble flow uniformity and stagnation characteristics are very important for the design of pebble bed high temperature gas-cooled reactor. Pebble flows inside some specifically designed contraction configurations of pebble bed are studied by discrete element method. The results show the characteristics of stagnation rates, recycling rates, radial distribution of pebble velocity and residence time. It is demonstrated clearly that the bed with a brachistochrone-shaped configuration achieves optimum levels of flow uniformity and recycling rate concentration, and almost no pebbles are stagnated in the bed. Moreover, the optimum choice among the arc-shaped bed configurations is demonstrated too. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity. In addition, a good design of the pebble bed configuration is suggested.

Effects of fuel particle size distributions on neutron transport in stochastic media

This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (keff) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron transport. As high as 1.00% relative difference in keff and ∼1.50% relative difference in peak fission power density are observed. As the packing fraction and optical thickness increase, the effect gradually dissipates. Neutron channeling between fuel particles is identified as the effect most responsible for the different neutronic results. Different size distributions result in the difference in the average number of fuel particles and their average size. As a result, different degrees of neutron channeling are produced. The size effect in realistic reactor unit cells is also studied and, from the predicted values of infinite multiplication factors, it is concluded that the fuel particle size distribution effects are not negligible.

Analysis of diffusion process and influence factors in the air ingress accident of the HTR-PM

Air ingress, one of the beyond design basis accidents for high temperature gas-cooled reactors, receives high attention during the design of the 250MW pebble-bed modular high temperature gas-cooled reactor (HTR-PM), because it may result in severe consequence including the corrosion of the fuel element and graphite reflector. The diffusion process and the set-up time of the stable natural convection after the double-ended guillotine break of the hot-gas duct are studied in the paper. On the basis of the preliminary design of the HTR-PM and its DLOCA analysis results, the diffusion process, as well as the influence of the core temperature distribution and the length of the hot-gas duct, is studied with the DIFFLOW code, which adopts a one-dimension variable cross-section diffusion model with fixed wall temperature. To preliminarily estimate the influence of chemical reaction between oxygen and graphite, which will change the gas component of the mixture, the diffusion processes between the He/N2, He/O2, He/CO and He/CO2 are calculated, respectively. Furthermore, the code has been improved and the varying wall temperature can be simulated. The more accurate analysis is carried out with the changing temperature distribution from the DLOCA calculation. The analysis shows that there is enough time to adopt appropriate mitigation measures to stop the air ingress and the severe consequence of fuel element damage and large release of fission product can be avoided.

Design of Gas-Cooled Fast Reactor 600MWth with Natural Uranium As Fuel Circle Input

This article presents the conceptual design of gas-cooled fast reactor (helium), the small size of the long-lived 600 MWth. Early stages of the design is to determine the geometry of the terrace, the value of the volume fraction and the mass fraction of fuel, cladding and coolant structure to calculate the parameters of reactivity, burnup, power distribution and density changes nuclides U238 and Pu239. The calculation is done using SRAC-CITATION code. SRAC code with JENDL-3.2 Data nuclides produced macroscopic cross section values for the eight energy group. Multi-group numerical solution of diffusion equations for 2-D geometry terrace RZ performed by CITATION code. The study results showed that the scheme Modified CANDLE, thermal power output is 600 MWth, with a fuel cycle for 10 years. This reactor has the advantage of requiring only the input of natural uranium in the fuel cycle, without the need for enrichment processes that affect the economic value. Keywords : Reactor, natural uranium, modified candle, burnup

ICONE19-43215 NUMERICAL RESEARCH ON THE THERMAL HYDRAULICS OF THE COOLANT IN A PEBBLE BED REACTOR CORE BY CFD

The water cooled pebble bed reactor with the pebble diameter from 2-10 mm has both the characteristics of high temperature gas-cooled pebble bed reactor and the traditional light water one. In the conceptual design of high temperature gas-cooled reactor and supercritical water-cooled reactor, pebble bed reactor has attracted much interest because of its outstanding advantages in the safety and fuel cycle. As far as china is concerned, it is worth researching and designing the water cooled pebble bed reactor, and the formers' researches were focused on the water cooled reactor. So it is great important to investigate the thermal-hydraulic characteristics of coolant in a pebble bed reactor core. The 3-D physical and geometrical models of the water cooled pebble bed reactor were built using the commercial software Pro/E and CFX, in which the thermal-hydraulic characteristics of the reactor core were studied. The velocity and temperature fields and pressure distribution of the coolant as well as the distribution of internal temperature of the fuel are obtained and analyzed in the case of diameter of 3mm and diameter of 6mm, and its effects on the core safety are also analyzed. The results indicate that the tendency of the same parameter values under different diameters are nearly identical, and it is larger with the diameter of 6mm, contrary, the flow disturbance with the diameter of 6mm is weaker than that of 3mm. The two maximum temperature of the pebble were found at the gap area. The maximum velocities of the first layer pebbles were found at the gap area, thereafter the maximum velocity was found between the gap areas. The pressure gradually reduces from bottom to top and the minimum values of pressure were generated at the gap area.

Similarity of high-temperature gas-cooled reactor technologies and designs in Russia and USA

Historical information concerning the development of high-temperature gas-cooled reactors in the USA and Russia is presented. The reactor facilities MHTGR (USA), VG-400 (Russia), VGM (Russia), GT-MGR (Russia, USA), and at the Fort St. Vrain nuclear power plant (USA) are described. The US programs for developing innovative high-temperature nuclear reactor technologies are examined. It is shown that the Russian and US technological developments for the fuel, reactor system, energy conversion system, and fission-product transport are similar.

Sensitivity study on depressurized LOFC accidents with failure of RCCS in a modular gas-cooled reactor

A modular gas-cooled reactor design with a thermal output of 600 MWt and a core exit temperature of 950 °C has been designed by the Korea Atomic Energy Research Institute based on the GT-MHR reactor concept which adopts a prismatic core. A sensitivity study on the transient plant behavior during a postulated depressurized LOFC accident concurrent with the failure of the RCCS was performed. In the transient analysis, the GAMMA+ code which can handle multi-dimensional, multicomponent problems was used. The RCCS is a passive system which is very reliable and supplies a significant heat removal mechanism during abnormal conditions in a GCR. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS with a depressurized LOFC was assumed. The thermal behavior of the reactor system was analyzed in various conditions. It is found that the maximum temperature of the reactor fuel compact could exceed 1600 °C at about 50 h at the condition of a depressurized LOFC with a failure of the RCCS. A problem with the structural integrity of the reactor pressure vessel could also be a critical factor. The insulation of a reactor cavity wall serves as a dominant obstacle against a heat transfer from the reactor vessel to the surrounding ground when the RCCS fails to operate. Without insulation material on the reactor cavity wall, the gradients of the increasing rate of the maximum temperature diminish and the peak values decrease. The maximum temperatures of the fuel compact and the reactor vessel are less sensitive to the concrete and surrounding soil properties, those are the thermal conductivity and volumetric heat capacity, when the insulation material is used. The uncertainties in the properties of the concrete and the surrounding soil become significant without an insulation material in the cavity. To improve the safety of a modular GCR, more effective and feasible heat removal mechanism need to be devised based on the comprehensions on the heat transfer characteristics.

ESTIMATION OF THE FISSION PRODUCTS, ACTINIDES AND TRITIUM OF HTR-10

Given the evolution of High-Temperature Gas-cooled Reactor (HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of 10 15 ~10 17 Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

Development of a method of evaluating an inventory of fission products for a pebble bed reactor

The design of the high temperature gas-cooled reactor (HTGR) has evolved and the relevant safety requirements have been defined; accordingly, the source term to be used as the basis for licensing must also be developed. However, analysis of the source term in the HTGR has not been adequately investigated and there has not been definite improvement in this respect. Because radioactivity in normal operation must be well understood, the purpose of this study is to establish a method for activity evaluation by the code combination MCNP-ORIGEN-MONTEBURNS-MOTEX. The sophisticated method, which constructs the HTR-10 core by using the unit lattice of a hexagonal prism, is developed for core modeling. The MCNP modeling is used to simulate the generation of fission products with an increase of burnup, and ORIGEN is utilized for depletion calculation of each fission product. Continuous fuel management is divided into five discrete periods for the feeding and discharging of fuel pebbles. MONTEBURNS is used for discrete fuel management. In short, this work by aid from MOTEX traces 41 isotope nuclides, the results of which seem highly probable. In addition, the inventory of actinides at the end of each cycle is also investigated. It would be informative when the waste management of spent fuel of HTGRs would be taken into account. This article lays the foundation for future work on the analysis of the source term in HTGRs and will hopefully serve as a platform from which the safety assessment of radioactive material release during accidents can be undertaken in future.