Reactor Site Research Articles

Study on an open fuel cycle of IVG.1M research reactor operating with LEU-fuel

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2–4 βeff for 20 years, and achieving a burnup of 9.9–10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.

Dose Distribution of Radioxenon Due to a Hypothetical Accident of TRIGA Research Reactor in Bangladesh

Radiological dose distribution owing to the deposition of 131m Xe, 133m Xe, 133 Xe, 135m Xe, 135 Xe, and 138 Xe on ground and immersion considering a postulated accident of TRIGA Mark-II research reactor has been assessed. The radiological dose distribution has been carried out in various directions with the help of Gaussian Diffusion Model. Local meteorological data such as average wind speed, frequency, etc. has been collected and evaluated for various directions around the reactor site. For all the dominant directions, the maximum dose values due to 131m Xe, 133m Xe, 133 Xe, 135m Xe, 135 Xe, 138 Xe and the total ( 131m Xe + 133m Xe + 133 Xe + 135m Xe + 135 Xe + 138 Xe) were observed within the limit 3.03E-7–1.23E-4 µSv/h, 1.01E-5–4.09E-3 µSv/h, 0.0003–0.14 µSv/h, 2.29E-5–9.26E-3 µSv/h, 0.002 –1.111 µSv/h, 1.11E-5–4.55E-3 µSv/h, and 0.003–1.269 µSv/h, respectively. Dose distribution was found to be dominant due to immersion and the contribution was 87.55 %. There is shortage of data regarding the release of radioxenon in the atmosphere during nuclear accident especially in the case of TRIGA type research reactor. This paper is the first such detailed study on atmospheric release of radioxenon and its dose distribution for a full power- reactor and the consequences towards the environment and public health. The result can be applied to develop the radiological protective measures and to prepare an emergency response plan for the TRIGA reactor site.

Enhancing the photocatalytic regeneration of nicotinamide cofactors with surface engineered plasmonic antenna-reactor system

A rational design of photocatalytic materials is essential to efficiently extract the photogenerated charge carriers, en route to achieving the desired chemical transformations. Here, we have incorporated the concept of ‘favorable catalyst-reactant interaction’ into the ‘antenna-reactor system’ to enhance the efficiency of charge carrier extraction in plasmonic photocatalysis. The benefits of coupling these two emerging strategies are utilized for the efficient visible-light photocatalyzed regeneration of nicotinamide cofactors. The photocatalytic antenna-reactor system comprises gold-rhodium nanoflowers (Au@Rh NFs), which are functionalized with negatively charged surface ligands to bring the favorable electrostatic interactions into action. The dual role of rhodium nanoparticles (RhNPs) – as an electron acceptor and as a reactor site – ensures an efficient extraction of the hot charge carriers from the plasmonic AuNP to the RhNP, and further to the reactants. Likewise, the introduction of favorable electrostatic interaction helps in channeling the electron mediators close to the catalyst surface, thereby enhancing the probability of the charge transfer process. Both these factors play a crucial role in the enhanced photocatalytic regeneration of nicotinamide cofactors by the Au@Rh NFs based antenna-reactor system. The photocatalytic activity of Au@Rh NFs based antenna-reactor system is well-retained for multiple cycles, proving their suitability for applied studies as well. Thus, our work showcases the power of coupling different concepts in plasmonic photocatalysis for enhancing the charge extraction process, which can be easily extended to different classes of solar to chemical fuel conversions.

The Design, Development and Assessment of a Novel De-centralised IoT-Based Remote Monitoring of a Small-Scale Anaerobic Digester Network

The implementation of community-level de-centralised anaerobic digestion (AD) systems offers a sustainable solution for organic waste management and energy provision, but there is currently a need for low-cost methods of system control and management at this scale. The problem becomes accentuated with increasing deployment, as greater complexities are created by the need to both monitor and control a wider network of smaller, community-scale plants. This paper describes research to design, deploy and test such a system by creating a network of two independent biogas generation AD reactor sites situated in the United Kingdom and Thailand. Internet of things (IoT) aspects such as inexpensive and widely available open-source control electronics was used in combination with small commercial gas analysers to log, transfer and share data with a central station, achieving real-time monitoring of performance for both networked digesters. The design proved beneficial for collaboration and knowledge exchange purposes along with providing off-site observations on reactor status. This information was used for early intervention to maintain biogas yields and enable the dynamic assessment of process economics. The concept showed potential for upscaling to larger multi-site networks of AD reactors that could maintain process optimisation without the need for skilled staff on site.

Low-threshold sapphire detector for rare event searches

Low mass nuclear recoil dark matter and coherent-elastic-neutrino-nucleus-scattering (CEνNS) searches confront similar challenges in choosing ultra-low threshold and large-mass detectors. We report experimental results from a 100 g single-crystal sapphire detector design with a diameter of 76 mm and thickness of 4 mm instrumented with transition edge sensors (TES). Sapphire is a crystal of aluminum oxide (Al2O3) and has been found to be a good candidate for light mass dark matter search experiments due to its lower atomic mass compared to other detector materials such as germanium and silicon. This new phonon-assisted sapphire detector was characterized to yield a baseline recoil energy resolution of 28.4 ± 0.4 eV. The detector is designed to be sensitive to low-energy rare interactions with an intention to investigate the low-mass region of dark matter phase-space and search for CEνNS at the reactor site.

Fission Battery transportation and siting aspects

Fission Batteries (FBs) are envisioned to be highly mobile, autonomously operating, and economically competitive small nuclear power reactor systems useable for generating electric power or heat. Such novel systems will face unique challenges pertaining to siting and transportation. This paper presents the findings and recommendations of the 2021 Idaho National Laboratory FB Initiative workshop on the transportation and siting of FBs. Regulatory and technical subject matter experts on the transportation and storage of radioactive materials, licensing of nuclear power plants, innovative sensing technologies, etc., participated in this workshop and shared their professional experiences and perspectives. Based on these conversations, the paper describes the lessons learned from nuclear reactor siting and licensing, as well as from the siting and transportation of radioactive materials such as fresh or spent nuclear fuel. It then explores the applicability of these lessons and practices to FB siting and transportation. In addition, associated regulatory and technological challenges are identified, such as a changing FB source term, criticality safety aspects, neutron activation and embrittlement of structural components, site development, transportation uncertainties, the current regulatory framework, modeling and simulation uncertainties, and limited inspections. Additionally, research and development activities are suggested to address these gaps and challenges, including the development of modeling and simulation tools for FB design, strategies for rapid FB transportation, innovative FB shell materials, procedures for onsite handling of FBs, transportation-accident-proof criticality control mechanisms, self-monitoring and self-testing capabilities, regulatory requirements, and probabilistic risk assessment tools. Carrying out these research initiatives is expected to help accelerate the evolution of FBs and facilitate their design, licensing, and deployment.

Siting of small modular reactors with renewable power generation support

Siting of small modular reactors with renewable power generation support

Assessment of radiological safety and emergency response of VVER-1200 type reactor

We studied the radiological safety and emergency response of the VVER-1200 type nuclear power plant using the RASCAL 4.3 and HOTSPOT 3.1.2 codes for INES levels 5, 6, and 7 reactor accidents due to SBO concomitant with LOCA for both dry and rainy seasons. The simulation was performed using Gaussian Plume Model and Lagrangian Gaussian Plume Model, respectively. The meteorological data for dry and rainy seasons were considered for the RNPP site of Bangladesh. The amount of total radioactivities of radiological equivalence to I-131 that released to the atmosphere for INES levels 5, 6, and 7 nuclear reactor accidents were 4011.6 TBq, 25,110 TBq, and 129,650 TBq, respectively. Radiological doses such as TEDE, TED, Thyroid CDE, Inhalation CEDE were evaluated for 40 km radial distance from the reactor site. The total effective dose for both RASCAL and HOTSPOT code predicts the higher dose value during the rainy season compared to the dry season. In the rainy seasons, TEDE and TED were very high about several ten times of permissible limit within 16 km radial distance. Therefore, immediate radiological protective measures are required to be implemented based on the emergency response plan of the local regulatory authority.

Preliminary Evaluation of Derived Concentration Guideline Level for Surface Soil at Wolsong NPP Site Using RESRAD-ONSITE Code

A preliminary derived concentration guideline level (DCGL) was calculated for the soil of the Wolsong nuclear power plant (NPP) (the first commercial pressurized heavy water reactor in South Korea) site using the RESRAD-ONSITE computational code. In total, fourteen selected radionuclides were analyzed after considering the preliminary evaluation information of radionuclides observed in the pressure tube specimen in the Wolsong Unit 1 heavy water reactor and previous NPP decommissioning cases. Furthermore, a geological structure model of the Wolsong NPP site was established according to the safety analysis report of the Wolsong NPP. In addition, the distribution coefficients (Kd) of various radionuclides were derived using the JAEA-SDB and the pH information of the groundwater around the Wolsong NPP site. The DCGL for surface soil of the Wolsong NPP site was derived via the application of criteria for the site release to facilitate unrestricted reuse. Moreover, preliminary dose evaluation and relevant analysis were performed according to the Wolsong NPP site resident scenario. The novelty of this study lies in the first calculation of the preliminary DCGL values for the case of the pressurized heavy water reactor (Wolsong NPP) site. It is expected that further reliable DCGL results might be achievable if more precise radionuclide information and site-specific parameters with respect to the Wolsong NPP site are secured and applied in the future.

A proposed approach for the evaluation of consequences of a large aircraft crash accident at an RBMK type reactor site during decommissioning

Nuclear power plants with RBMK reactors were designed in late sixties - early seventies and were built entirely in the European part of the Russian Federation, the Lithuanian Republic and the Ukrainian Republic. The plants in the Lithuanian and the Ukrainian Republics are in various stages of decommissioning. As currently planned, the most of RBMK reactors still operating in the Russian Federation will end their operation in this decade.The reactor building, spent nuclear fuel and radioactive waste storage facilities, built at the time of construction of RBMK plants are not specially designed to withstand the direct impact of an aircraft crash. There were no such regulatory requirements at the time the plants were designed and, no such requirements were introduced by regulatory bodies later on. But for now, in the aftermath of the September 11, 2001 terrorist attack on the United States, the event of an aircraft crash on a nuclear facility can no longer be considered as extremely unlikely.This paper considers the state of a nuclear power plant with the RBMK type reactor(s) during certain decommissioning and dismantling stages and discusses the approach for the evaluation of radiological consequences to the public in the event of a large aircraft crash on potentially hazardous objects of the plant. The damage to spent nuclear storage facilities, the reactor graphite stack and storage facilities for the combustible waste may cause a significant release of radionuclides to the environment. Key components of the approach include the evaluation of a structural damage to the facilities, the evaluation of a fire propagation and impact, the evaluation of a spent nuclear fuel cooling capability in storage pools, the evaluation of an airborne radionuclide release, the evaluation of a radionuclide dispersion in the environment and the evaluation of the public exposure. A set of widely used software and methods for the performance of the specific tasks is discussed. The paper also describes how the proposed approach was applied for the evaluation of radiological consequences from a Boeing 767-400 type aircraft crash at the Ignalina NPP site where a plant with two RBMK-1500 type reactors is decommissioned.The paper does not address modern radioactive waste and spent nuclear fuel management and storage facilities that might be built at RBMK sites as well. As a rule, assessment and mitigation of an aircraft crash accident consequences is an integral part of design of these new facilities.

Validation of UNF-ST&DARDS As-loaded safety analysis methods for BWR decay heat calculations

A novel assessment of the conservatism in the UNF-ST&DARDS decay heat calculations has been performed. UNF-ST&DARDS is used to quantify the uncredited margin in safety analysis calculations for spent nuclear fuel (SNF) storage, transportation, and disposal systems. The goal of the assembly-specific as-loaded safety analysis approach in UNF-ST&DARDS is to determine the time-dependent realistic state of the SNF systems; however, it is desirable to conservatively estimate safety analysis parameters, such as decay heat, for a given set of fuel characteristics. The primary source of conservatism in the generic UNF-ST&DARDS assembly-specific as-loaded analysis (also referred to as bounding within UNF-ST&DARDS) is the conservative assumptions of various reactor operational parameters that attempt to envelop wide spectrum of reactor operating scenarios. The assessment in this paper is necessary to demonstrate that sufficient decay heat conservatism is retained in the UNF-ST&DARDS bounding as-loaded spent fuel analysis methodology. This paper also demonstrates the time dependent impact of various parameters such as last cycle power on decay heat values. A comparison between the UNF-ST&DARDS bounding decay heat calculations and calculations performed using a detailed description of the fuel assembly operating histories, referred to as detailed calculations, was performed using recently acquired data. The data used to perform this evaluation are from one set of 3019 assemblies from a US boiling water reactor (BWR) site and one set of 2117 assemblies (952 8 × 8, and 1165 10 × 10) from a Swedish BWR reactor. Analyses of the US data involved two sets of assumptions for the bounding calculations and produced two data sets. The first analysis, in which the cycle-wise burnups were derived for the bounding calculations from the detailed data, generated the derived data set; the second, in which the assumptions associated with incorporating US nuclear fuel data survey (Form GC-859) data were included in the calculations for a subset of the same assemblies, generated the GC-859 data set . When bounding assumptions were used, the average level of conservatism (overestimation of decay heat) ranges between 9.0% and 17.7% for the derived data set, between 11.4% and 32.3% for the GC-859 data set, between 10.1% and 62.6% for the Swedish 8 × 8 fuel and between 8.3% and 44.7% for the Swedish 10 × 10 fuel. The level of conservatism and the scatter in the ratio between bounding and detailed data increase significantly for the 100- and 200-year cases for derived US and Swedish data sets. The GC-859 data set had large conservatisms in the early cooling times that initially shrank with time and then increased for the 100-year and 200-year cooling times. These results show that, while UNF-ST&DARDS can be used to calculate assembly decay heat based on assembly characteristics and operating history to identify potential significant margins to the licensing basis decay heat calculations, the decay heats calculated by UNF-ST&DARDS are still conservative compared to more detailed calculations for the range of assemblies and operating conditions assumptions evaluated in this study.

An integrated method for atmospheric dispersion and corresponding risks: Application to ITU TRIGA Mark II research reactor

The prediction of the present and future impact of radioactivity release is important. However, there are several models and options to choose from. An integrated approach is proposed in this study and it is implemented to Istanbul Technical University (ITU) TRIGA Mark II Research Reactor. The procedure was divided into six steps including quantification of the source term, specification of meteorological data, atmospheric dispersion model selection, atmospheric dispersion and deposition calculation, biosphere and geosphere transport calculation, and finally dose and risk calculation. The acquisition of data and parameters for these steps were discussed in detail to provide insight to users of the integrated method. The evaluation of wind direction and speed defined in the safety analysis report of the reactor showed that the average wind direction at the reactor site altered from NE direction to NE-NNE over time but speed stayed almost the same. The maximum-total cloudshine and groundshine as a result of annual exposure rate were found as 21.1 μSv/yr at ITU Ayazağa Campus. For inhalation, Ba-140 and I-131 were found to have the highest risk in mortality and morbidity, respectively. The same is true for Sr-90 and I-131 in the lifetime intake of food while in the lifetime intake of tap water, Sr-90 is the leading radioisotope for both mortality and morbidity risks. However, all calculated values are quite low considering the survival rates. Moreover, the results of the study provides useful insight into the understanding of necessities while shedding light on limitations.

First results from a search for coherent elastic neutrino-nucleus scattering at a reactor site

The deployment of a low-noise 3 kg p-type point contact germanium detector at the Dresden-II power reactor, 8 meters from its 2.96 GW$_{th}$ core, is described. This location provides an unprecedented (anti)neutrino flux of 8.1$\times 10^{13} ~\bar{\nu_{e}}/$cm$^{2}$s. When combined with the 0.2 keV$_{ee}$ detector threshold achieved, a first measurement of CE$\nu$NS from a reactor source appears to be within reach. We report on the characterization and abatement of backgrounds during initial runs, deriving improved limits on extensions of the Standard Model involving a light vector mediator, from preliminary data.

Development of an optimization-aided small modular reactor siting model – A case study of Saskatchewan, Canada

As an emerging clean technology, the replacement of coal plants and the integration with renewable power generation are two appealing options for the deployment of SMRs in the power system. However, both these applications and SMR siting issues for power systems have multiple complexities that have been difficult to be considered in past studies. Accordingly, an optimization-aided small modular reactor siting (OASS) model is initiated in this study to support decisions of siting, sizing, and timing for small modular reactor (SMR) development. The proposed OASS model is specialized for obtaining possible SMRs siting schemes, wind-SMRs combination schemes, and power system transition schemes with coal plant closure and SMRs deployment. This research initiates the application of an optimization-based modelling approach to tackle SMRs site selection issues for Saskatchewan, Canada. The results disclose that electricity imports and natural gas power will fill the electricity demand gap in Saskatchewan’s power system in the short term. Results obtained from this model can also identify optimized patterns of SMRs and wind farms deployment and siting in Saskatchewan. Power stations with large capacity and independent transmission grid will become the primary choice in the SMRs and wind farms site selection process. Such patterns and alternatives would enable decision-makers obtain optimized energy structure by introducing SMR reflecting trade-offs between overall system costs and greenhouse gas (GHG) emissions from both economic and environmental perspectives.

Preprocessing Energy Intervals on Spectrum for Real-Time Radionuclide Identification

In this study, we present a preprocess method using radiation energy intervals on a gamma-ray spectrum based on a deep learning algorithm to achieve real-time radionuclide identification. Data preprocessing is performed by classifying energy intervals, distinctly corresponding to pulse amplitudes of each radiation measurement system. Since the energy intervals are distinguished with noise, backscatter area, Compton edge, and photopeaks depending on radionuclides, raw data are sorted in each interval in preprocessed dataset using a deep learning algorithm. Using <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">60</sup> Co, <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">137</sup> Cs, and the energy interval preprocessing, the multi-source identification shows 100% accuracy in 2000 measured data compared with 70% accuracy for those without the preprocessing method. The measured time is 72 s for 2000 test data, dramatically reduced from the conventional data collection time of 60 min for 100 000 data. The proposed approach reduces the minimum number of data to identify radionuclides before visualizing the spectrum. With the preprocess method, radionuclide identification is completed in tens of seconds, applicable for low radiation activity areas such as decommissioning reactor sites.

СИСТЕМА АВТОМАТИЗИРОВАННОГО ПРОЕКТИРОВАНИЯ И ПРОТОТИПИРОВАНИЯ УЧАСТКОВ СИНТЕЗА УГЛЕРОДНЫХ НАНОКЛАСТЕРНЫХ СТРУКТУР

The article focuses on using the integrated multi-component computer-aided design and control systems, which changes the product life cycle processes in the high-tech complex chemical and technological systems. In the integrated design of production facilities, special attention is paid to safety, productivity, reliability, ergonomics of the production line and the material and technical environment. The most effective in project management are flexible life cycle models that connect diverse specialists in a single information environment. The relevance of the task of developing and implementing problem-oriented heterogeneous computer-aided design and digital prototyping systems is justified. There is described a method for improving the efficiency of designing production facilities while minimizing the time required preparing a design solution, and the risks of design errors caused by the communication dissonance of diverse specialists. The architecture of the computer-aided design and digital prototyping system with a description of its basic components is presented. The creation of a digital model of the design object is described on the example of the reactor site and the material and technical base of the process of synthesis of carbon nanocluster structures that are in high demand in various industries. The conducted tests and the introduction of the resulting applied software solution into the educational process proved its efficiency in solving problems of integrated computer-aided design and digital modeling not only of objects of the carbon nanoindustry, but also for other industries with a similar material and technical base.

Human factors engineering and participatory design in the project of a nuclear research reactor control centre

The Human Factors Engineering (HFE) program is an essential aspect for the design of nuclear installations. The overall aim of the HFE program is the improvement of the operational reliability and safety of plant operation. The HFE program main purpose is to ensure that human factor practices are incorporated into the plant design, emphasizing man-machine interface issues and design improvement of the nuclear reactor Control Centre. The Control Centre of nuclear reactor is a combination of control rooms, control suites and local control stations, which are functionally connected and located on the reactor site. The objective of this paper is to present a participatory design approach, including human factor issues, for the Control Centre of a nuclear research reactor used to produce radioisotopes and for nuclear research. The design approach is based on participatory design principles, using human factor standards, ergonomic guidelines, and the participation of a multidisciplinary team during all design phases. Using the information gathered, an initial sketch 3D of the Control Centre was developed.

Bringing Earth-Abundant Plasmonic Catalysis to Light: Gram-Scale Mechanochemical Synthesis and Tuning of Activity by Dual Excitation of Antenna and Reactor Sites

The localized surface plasmon resonance (LSPR) excitation in plasmonic nanoparticles (NPs) in the visible and near-infrared ranges is currently at the forefront of improving photocatalytic performances via plasmonic photocatalysis. One bottleneck of this field is that the NPs that often display the best optical properties in the visible and near-infrared ranges are based on expensive noble metals such as silver (Ag) and gold (Au). While earth-abundant plasmonic materials have been proposed together with catalytic metals in antenna–reactor systems, their performances remain limited by their optical properties. Importantly, the synthesis of plasmonic photocatalysts remains challenging in terms of scalability while often requiring several steps, high temperatures, and special conditions. Herein, we address these challenges by developing a one-pot, gram-scale, room-temperature synthesis of earth-abundant plasmonic photocatalysts while improving their activities beyond what has been dictated by the LSPR excitation of the plasmonic component. We describe the mechanochemical synthesis of earth-abundant plasmonic photocatalysts by using MoO3 (antenna) and Au (reactor) NPs as a proof-of-concept example and demonstrate that the dual plasmonic excitation of antenna and reactor sites enables the tuning of plasmonic photocatalytic performances toward the reductive coupling of nitrobenzene to azobenzene as a model reaction. In addition to providing a pathway to the facile and gram-scale synthesis of plasmonic photocatalysts, the results reported herein may open pathways to improved activities in plasmonic catalysis.

Statistical analysis of extreme value of meteorological elements observed for the last 31 years (1989–2019) at Narora site

Understanding of extreme weather conditions at the site of interest is essentially required in the design of engineering structures so that the structures can withstand weather stresses. This paper presents an analysis of extreme values of meteorological elements observed at Narora site for the last three decades (1989–2019). The elements considered are extreme air temperature (°C), minimum relative humidity (%), extreme wind gust (km/h), maximum rainfall (mm) in a day and a month, and annual rainfall. The extreme value analysis reveals that the maximum air temperature, maximum wind gust at 30 m, maximum monthly rainfall, and maximum annual rainfall obey Fisher-Tippett Type-1/Gumbel extreme value distribution, whereas minimum air temperature, minimum relative humidity (%), annual daily maximum rainfall (mm), and annual minimum rainfall (mm) obey Fisher-Tippett Type-2/Frechet extreme value distribution function. Distribution function parameters, i.e., location, scale, and shape parameter for each variable, have been determined. Extreme values corresponding to return periods of 50, 100, and 1000 years are worked out using best fit linear regression curve as a compliance of the Atomic Energy Regulatory Board Safety Guide Recommendations. The derived extreme values are particularly useful to designer for arriving at suitable design basis values of different elements to ensure the safety of the reactors and other civil structures in Narora region, with respect to stresses due to weather conditions. Extreme values corresponding to return periods of 50 and 100 years at Narora are compared with corresponding values at other three nuclear reactor sites in India, namely, Tarapur, Kalpakkam, and Trombay. In addition, the time series pattern analysis of rainfall for 31 years at the Narora site closely following the 2-year moving average rainfall data pattern. These results can be used for water harvesting, irrigation, and floods management plans in the future.

A chance-constrained small modular reactor siting model -- a case study for the Province of Saskatchewan, Canada

In this study, a chance-constrained small modular reactor siting (CCSS) model has been first proposed for planning regional electric power systems in the Province of Saskatchewan under the pressure of greenhouse gas emission mitigation. Through incorporating three programming methods (interval linear programming, chance-constrained programming, and mixed-integer programming) in an optimization framework, the CCSS model could effectively deal with multiple uncertainties expressed as probability distribution and intervals in constraints and objectives. Since locations of retired coal-fired power stations have been proposed as new sites for small modular reactor (SMR), this model could provide the construction planning for SMR in terms of its sizes and sites. Various solutions to power generation and capacity expansion with different risk levels were obtained under the objective of minimal system costs. Results are helpful for identifying optimized strategies to increase the proportion of renewable energy under emission constraints. In addition, model without introducing the technology of SMR has been comprehensively compared with the CCSS model to recognise the effects of SMR on system costs and greenhouse gas emissions. Results indicate that the CCSS model is effective for supporting long-term clean and renewable energy utilization as well as emission reduction policy formulation in Saskatchewan.