Nuclear Safety Research Articles

How useful is setting safety goals?

The study explores the multifaceted role of safety goals in fostering a risk-informed culture, reflecting the global experience within the nuclear safety domain. Analysis of the phases of transition evoked by establishing and applying safety goals sheds light on the need for epistemic, reflexive, and practical transitions for better management of nuclear safety. This pivotal role of safety goals underscores the importance of recognizing them not as ready-to-use turnkey products but as catalysts for stakeholder dialog, reassessment of existing safety paradigms, and regulatory framework refinement. Finally, this study explores the challenges associated with standardizing safety goals globally and navigating the transition process within the framework of transition management.

Enabling pulse shape discrimination with commercial ASICs

Fast electronic readout for high-channel density scintillator-based systems is needed for radiation tracking and imaging in a wide range of applications, including nuclear physics, nuclear security and nonproliferation. Programmable electronics, like FPGAs and ASICs, provide a fast way of conditioning and processing the signal in real time. In this paper, we present a pulse shape discrimination (PSD) method based on the shaping circuit of a commercially available ASIC, the Citiroc1A by CAEN Technologies. We used two different shaping times per detector channel to calculate a shaping parameter that enables PSD. Using our new method, neutron and gamma-ray pulses detected by a d12-stilbene scintillator can be effectively discriminated at light output values greater than 0.15MeVee. While not achieving the PSD performance of traditional offline charge integration, our method does not require the transfer of data to a separate system for further processing and enables the direct deployment of high-channel density multi-particle detection systems. Moreover, the availability of a wider range of shaping times than those on the Citiroc1A can potentially further improve the PSD performance.

Legal regulation of nuclear and radiation safety during operation of nuclear power plants

The article examines the problems of legal regulation of nuclear and radiation safety that arise during the operation of nuclear power plants. On the basis of the legal definition of such legal categories as radiation and nuclear safety, radiation protection, which must be guaranteed by the state to human life and health, the main attention is focused on the analysis of the legal regime of lands as the basis of human life activities that have undergone radioactive contamination, which are legally classified as man-made polluted. The genesis of the legislation that secures the definition of these lands is analyzed, it is emphasized that these lands are the result of human economic activity, that is, it is about the consequences of anthropogenic activity, when highlighting the peculiarities of the legal regime of man-made contaminated lands, their division into radioactively dangerous and radioactively contaminated is taken as a basis. The mentioned lands, being in state ownership, cannot be transferred to private ownership, are included in the exclusion zone and the zone of mandatory (unconditional) resettlement of the territory that has undergone radioactive contamination as a result of the Chernobyl disaster. Residence and permanent stay of people in the territory of these zones is prohibited by law. Land plots located in the zone of guaranteed voluntary resettlement are also radioactively contaminated and are used in accordance with the procedure determined by the Cabinet of Ministers of Ukraine. The amendments to the current legislation are analyzed separately, which relate to the inclusion of radioactively contaminated lands in the ecological network for the purpose of restoration and are related to the creation of the Chernobyl radiation- ecological biosphere reserve. It is emphasized that part of the territory of the reserve, created at the expense of the zone of exclusion and the zone of unconditional (mandatory) resettlement, being on the border with the Republic of Belarus, causes additional restrictions of its legal regime. In this case, we are talking about changing the purpose of a part of the relevant lands, which have been transferred to defense lands. The conclusions emphasize that lands that have undergone radioactive contamination need priority restoration, which today is carried out naturally. The legal mechanism for restoring the state of such lands is imperfect and needs significant changes. It is time to develop and approve the Strategy for returning radioactively contaminated land to economic use.

The structural, optical, electrical and radiation shielding properties of Co-doped ZnO thin films

Co:ZnO thin films are promising due to the similar ionic radii of cobalt and ZnO, and their potential for carrier spin injection, transfer, and detection. This study aims to investigate the structural, optical, electrical, and radiation shielding properties of Co-doped zinc oxide (Co:ZnO) thin films for potential applications in various fields such as semiconductors, solar cells, spintronics, and radiation shielding. The effect of commercially available Cobalt acetate tetrahydrate and Cobalt complex (Co with conjugated NN ligand) as cobalt additives on the synthesis of ZnO thin films is compared, aiming to discern their respective impacts. The dopants are incorporated in 2, 4, and 6 wt%, employing the sol-gel spin coating method.Applying cobalt complex additive as Co precursor resulted in ZnO:Co films with a superior enhancement of the electronic band gap of 2.84–3.00 eV compared to cobalt acetate precursor (2.62–2.84 eV). With the introduction of Cobalt, there is a notable change in the electrical conductivity, showcasing a decrease from 1.5x10−7 to 2.11x10−10 (S/cm). Also, radiation shielding parameters were experimentally measured in the energy range of 186.1–1764.5 keV within the scope of gamma ray spectral analysis. Exposure buildup factors (EBF) in the energy range of 15–15,000 keV were calculated up to 1–40 mean free path (mfp). Mass stopping power (MSP) and projected range values (PR) for nuclear safety applications were calculated for protons and alpha particles using the SRIM code. Thin film thicknesses were determined by using Lambert-Beer law. It was observed that the thin film thicknesses obtained from gamma spectrometry (2.090–2.628 μm) and SEM (2.157–2.725 μm) were compatible with each other.

Sulfur vacancy-rich bismuth sulfide nanowire derived from CAU-17 for radioactive iodine capture in complex environments: Performance and intrinsic mechanisms

Effective capture and immobilization of volatile radioiodine from the off-gas of post-treatment plants is crucial for nuclear safety and public health, considering its long half-life, high toxicity, and environmental mobility. Herein, sulfur vacancy-rich Vs-Bi2S3@C nanocomposites were systematically synthesized via a one-step solvothermal vulcanization of CAU-17 precursor. Batch adsorption experiments demonstrated that the as-synthesized materials exhibited superior iodine adsorption capacity (1505.8 mg g−1 at 200 °C), fast equilibrium time (60 min), and high chemisorption ratio (91.7%), which might benefit from the nanowire structure and abundant sulfur vacancies of Bi2S3. Furthermore, Vs-Bi2S3@C composites exhibited excellent iodine capture performance in complex environments (high temperatures, high humidity and radiation exposure). Mechanistic investigations revealed that the I2 capture by fabricated materials primarily involved the chemical adsorption between Bi2S3 and I2 to form BiI3, and the interaction of I2 with electrons provided by sulfur vacancies to form polyiodide anions (I3-). The post-adsorbed iodine samples were successfully immobilized into commercial glass fractions in a stable form (BixOyI), exhibiting a normalized iodine leaching rate of 3.81 × 10−5 g m−2 d−1. Overall, our work offers a novel strategy for the design of adsorbent materials tailed for efficient capture and immobilization of volatile radioiodine.

Survey Analysis of Potential Nuclear Safety Research of Thailand for International Research Collaborative Reinforcement in the 2020s

Abstract To achieve the long-term challenge of nuclear energy public acceptance in Thailand, nuclear safety research needed to be properly determined in both domestic and international directions, especially in the 2020s which was a period passing the Fukushima disaster over 10 years. Thailand Institute of Nuclear Technology (TINT) has studied nuclear safety research after the Fukushima accident to answer technical and social issues of nuclear power. An update of nuclear safety research from domestic experts and international surveys was needed in order to identify potential collaborative research to serve the goal of public acceptance reinforcement. The objective of this study was to survey, assess and rank the importance and knowledge level of nuclear safety research in Thailand among domestic experts in various fields. The survey was extended to collect the opinion of international participants of the ASEAN Network on Nuclear Power Safety Research (ASEAN NPSR) to analyze the similarity of the nuclear research interest for reinforcing the future collaborative project. As a result, the importance and knowledge level showed diverse important research topics with the priority of research scopes on human factor novel reactor technologies, and risk assessment. According to the ASEAN NPSR survey, the nuclear safety research of severe accidents, risk assessment, and novel reactor technologies were listed as potential collaborative projects. Also, the domestic and ASEAN NPSR survey results helped support the new collaborative research extension session in the annual ASEAN NPSR meeting to together discuss the potential nuclear safety research between members for the 2020s.

Basal and prismatic slip in hafnium

Metals with hexagonal crystal structure are found in many safety-critical applications, from titanium alloy fan blades in jet engines to magnesium alloys in the automotive industry. They also exist in nuclear reactors, for example hafnium alloy control rods whose structural integrity is critical to nuclear safety. However, mechanistic understanding of the micromechanical properties and deformation behaviour of Hf remains elusive. To aid in this understanding, we performed in situ scanning electron microscopy single crystal micropillar compression tests of a commercial Hf alloy, where the samples were aligned to activate 〈a〉 prismatic and 〈a〉 basal slip systems. We then employed transmission electron microscopy to examine the dislocation structure in the deformed pillars. These two slip systems exhibited distinctly different behaviour, where prismatic slip is planar and basal slip is wavy. Prismatic slip is the easiest deformation mode, while basal slip is at least twice as difficult. Generally, the characters of prismatic and basal slip in Hf show similarities to those in some other hexagonal metals such as Ti and Zr, except for the higher relative difficulty of basal slip. Our results provide fundamental knowledge and parameters that can be used for modelling degradation and predicting performance of Hf, and shed light on the general behaviour of basal and prismatic slip systems in low-c/a-ratio hexagonal metals.

Improvement of butt-fusion welding procedure and performance evaluation method for thick-walled bimodal polyethylene pipes

High-density polyethylene (HDPE) pipes have been continuously improved by the progress made by the resin industry and pipe manufacturers. Together with the advanced technology of welding equipment in terms of structure reliability and automations, HDPE pipes have been gradually applied to higher pressure level, bigger diameter, and thicker walls in operations which have become the preferred piping system for nuclear safety application in transporting cooling seawater. There are substantial differences in parameter settings when applying heat and force during different welding procedures, as well as the complicated evaluation methods for mechanical properties, which causes communication barriers between engineers and researchers. Thus, it is necessary to comprehensively study the welding procedures and evaluation methods to standardize the evaluation protocols in the industry. In this paper, based on the assumptions of temperature effect and molecular chain diffusion dynamics at the welding interface, the optimization of the welding procedure is proposed. Some well-established evaluation methods have been used for performance assessment and their feasibilities are discussed. The idea of creating crack along the weld interface is adopted as new method to evaluate the material ability of fracture resistance. Specimens with unilateral notch and weld interface crack are proposed and subjected for quasi-static axial tensile test. The fracture stress and fracture energy at the welding interface are used to differentiate the quality of the joint from different welding procedures. The fracture morphologies of the weld interface are analyzed. A complete picture is drawn for the mechanical fracture resistance of the weld interface. The method used in this paper can be corelated to the pipeline design conditions, therefore could be applied to the real-time engineering applications. Moreover, the evaluation method proposed in this paper is simple, reliable, and easy to implement.

Specific Grammatical Aspects of Translating the Iaea Regulatory Documents into the Russian Language

The article discusses some grammatical features of the IAEA safety requirements and guides in the English language and in their Russian translations. Considering that these documents make a foundation of the global nuclear safety system, and grammar is the backbone of the language and an essential constituent of the communication content, the paper is focused on the analysis of the ways of translating specific grammatical structures, which determines current importance of this study. The aim of the work is to compare typical grammatical constructions used in the texts of the Russian translations of the IAEA publications under consideration with the grammatical features of similar Rostechnadzor’s regulatory documentation. The main tasks of the authors are to compare the ways of translating English modal and passive constructions, English verbals and verbal structures into Russian and to verify the translator’s choice of a certain grammatical structure in the translations of the regulations and guidelines under consideration. Taking research tasks predetermined the choice of the following research methods: comparative analysis; linguistic and statistical analysis; contextual analysis; inductive and deductive analysis. To obtain linguistic and statistical analysis data, the authors applied such software tools as SDL Trados, Adobe Acrobat, Excel and MS Word. In order to ensure the correctness of the results, only the substantive part of the documents under consideration was analyzed.The analysis has brought to the following results: the need to comply with the regulatory requirements narrows the choice of Russian modal constructions in the translation of "should" statements used in IAEA documents. The issue of changing the regulatory requirement that prescribes to use the Russian verb "следует" has to be treated very carefully, considering the established practice of both translating the IAEA safety guides and their understanding by the target users; the translation of English passive constructions and verbals used in the IAEA safety guides gives clear preference to the use of Russian indefinite-personal structures with verbs in the active voice in the 3rd person plural; the grammatical structures used in the translation of the IAEA publications in question generally correspond to the grammatical structures used in the Rostechnadzor regulatory documents, which is predetermined by the information nature and ensures a fairly high similarity of their style.

Modified Evaporation Model for ATHLET System Code in a Passive Containment Cooling System for Nuclear Safety

The presented work deals with the improvement of the evaporation model of the ATHLET (Analysis of Thermal and Hydraulics of Leaks and Transients) system code to be applied to a passive containment cooling system of a nuclear power plant. For the model validation, INTRAVIT (Investigation of Passive Heat Transfer in a Variably Inclined Tube) test facility setup at the University of Luxembourg was used. The first part of the paper presents a review of the existing literature on evaporation models that revealed that those models significantly simplify the physical processes that occur. Next, a modified evaporation model is proposed that offers a realistic description of various evaporation processes and the start of bubble formation using a nucleation model, and a surface density calculation model is introduced that is necessary for evaporation simulation. The final part of this work explored five different system configurations to test the evaporation model: three condenser tube inclinations (5 deg, 60 deg, and 90 deg), two riser lengths (1 m and 2.5 m), and different thermal loads. They made it possible to simulate several experiments for stable and unstable natural circulation and to verify the proposed model.

Numerical Investigation of Process Parameter Effects on Flow Oscillations in a Natural Circulation Boiling Integral PWR-Type SMR Test Rig

Innovative reactor designs like small modular reactors (SMRs) have the potential to operate in a natural circulation (NC) boiling mode, but this mode introduces flow oscillations that pose a risk to nuclear safety. Therefore, it is essential to investigate the effects of various parameters on these oscillations. This study focuses on predicting the operational behavior of the Integral PWR-type SMR Test Rig (iPSTR) when operating in NC and subcooled boiling conditions. The iPSTR replicates an NC boiling loop with a vertical heater, vertical cooler configuration, high-temperature and high-pressure conditions, and nonuniform diameter structure. Using the RELAP5 model, thermal-hydraulic simulations were performed to anticipate how varying degrees of inlet subcooling affects parameters such as mass flow rate and void fraction, with experimental data used to validate the model’s accuracy. This investigation covers a range of process conditions, including system pressures from 5 to 20 bars, core input power varying from 8.5 to 14.5 kW, and degrees of inlet subcooling from 1 to 49 K. The results reveal that increasing input power leads to higher average mass flow rates, while at a constant system pressure, higher input power stabilizes flow rates at higher degrees of inlet subcooling. Moreover, reduced and more consistent oscillation amplitudes and frequencies at higher core power result at more elevated system pressure, enhancing the safety of the iPSTR facility.

A sensitivity study on the PDFs treating uncertainties in severe accidents for pressurized heavy water reactors

This research article introduces a study regarding the uncertainties treatment during severe accidents for Pressurized Heavy Water Reactors (PHWRs). The present study is focused upon the unmitigated Station BlackOut (SBO) accident analysis for a CANada Deuterium Uranium (CANDU) type reactor emphasizing the impact of the uncertainties treatment on the relevant key timings of the SBO accident progression through different approaches for the uncertainty parameters’ Probabilistic Distribution Functions (PDFs). A comparison between the sensitivity analysis results is provided in the present research study.The uncertainty analysis is performed with the RELAP/SCDAPSIM code with the Integrated Uncertainty Analysis (IUA) package from the code.Results from the research would support the advancements on the best-practices for uncertainty analyses with respect to the parameter's uncertainties distribution functions. Data dispersion is a key element for the realistic quantification of uncertainties in nuclear power plant safety analyses, including severe accidents.

Comparative study on the performance of butt fusion-welding processes for nuclear safety class large-diameter thick-walled PE pipes

New technologies in polymer synthesis and pipe extrusion equipment have led to the commercialization of high-performance, large-diameter, thick-wall high density polyethylene (HDPE) pipes. They have been used in the field of seawater transport and cooling to replace metal pipes, due to their advantages of high corrosion resistance and extensibility. Connection of HDPE pipe is important as it determines the safety of the entire piping system. Butt fusion welding is commonly used for HDPE pipe connection but may cause the formation of weak points in the welded joints, interfering the reliability of the pipeline system in the application of nuclear power plants. At present, there is a lack of research on evaluating the performance of welded joint for large-diameter thick-wall HDPE pipes made by butt fusion-welding. The purpose of this study is to investigate the influence of three different butt fusion-welding processes, i.e., single low pressure (SLP), single high pressure (SHP) and dual low pressure (DLP), by evaluating the performance of their welded joints, including characterizing tensile strength, extensibility, crystallinity and hardness. In specific, a thick-wall HDPE pipe with OD of 812.8 mm and wall thickness of 74 mm which is certified for nuclear safety class was used for study. Representative specimen from the outer, middle and inner part across the wall of the main pipe body and welded joints were taken for testing. Different test methods and specimens were designed to assess the feasibility of evaluating the welding performance from different welding process. The results showed that the mechanical properties of different locations of the welded joints were different, and the tensile strength and fracture energy of the middle part of the joint were lower than that of the inner and outer parts, which could be caused by the difference in the crystallinity and thickness of the melting zone influenced by welding processes, as can be seen from the analysis of DSC test and morphology observation. Hardness testing was conducted on the section of the welded joints, and it revealed that the micromechanical properties of the welded joints in the region of the heat-affected zone were enhanced significantly, which may be due to the annealing effect caused by welding process. In summary, The DLP process resulted in the best extensibility of the welded joints among three processes, suggesting that the joining pressure from welding process plays an important role in affecting the extensibility of the welded joints.

US nuclear agency struggles with production and costs

The National Nuclear Security Administration must cope with resurgent geopolitical threats accentuated by Vladimir Putin’s nuclear saber-rattling.

Hydrophobic nanosheet silicalite-1 zeolite for iodine and methyl iodide capture

Effective capture of radioactive iodine from nuclear fuel reprocessing is of great importance for public safety as well as the secure utility of nuclear energy. In this work, a hydrophobic nanosheet silicalite-1 (NSL-1) zeolite with an adjustable size was developed for efficient iodine (I2) and methyl iodide (CH3I) adsorption. The optimized all-silica zeolite NSL-1 exhibits an excellent I2 uptake capacity of 553 mg/g within 45 min and a CH3I uptake capacity of 262 mg/g within 1 h. Benefiting from the reduced thickness and enhanced porosity, microporous NSL-1 possesses enhanced iodine adsorption capacity and fast adsorption kinetics, which is a considerable high value among inorganic materials. Unexpectedly, the remarkable characters of high hydrophobicity, acid-resistance and anti-oxidation endow it a higher iodine uptake capacity than traditional aluminosilicate zeolites. More importantly, the high uptake selectivity toward I2 possessed by NSL-1 owing to its hydrophobic skeleton under simulated dynamic conditions. The low cost, facile and scalable synthesis of NSL-1 further highlights great prospects for applications in the nuclear industry. This work provides useful insights for designing efficient adsorbents for iodine capture.

Timing bombs and the temporal dynamics of Iranian nuclear security

Abstract For more than two decades, Iran’s nuclear programme has concerned policymakers and scholars alike. Whether speeding up uranium enrichment, slowing down international negotiations, or disrupting the timing of key initiatives, actions around Iran’s nuclear programme bear a clear time signature. Yet systematic accounts of the importance of time in shaping foreign and security policymaking have been largely neglected. Through foreign policy timing theory’s (FP4D) reconceptualisation of time we show how actors both constructed and then used time to pursue their strategic interests, creating, altering, and sabotaging the timing mechanism linking Iranian nuclear technology and international sanctions. These manipulations of time by both domestic and international actors resulted in prolonged international negotiations and fluctuating periods of crisis and produced a temporally flawed agreement frozen in time. We consider time’s impact on the current challenges and future direction of nuclear diplomacy with Iran as well as its importance for broader nuclear security issues.

Keep your enemies safer: technical cooperation and transferring nuclear safety and security technologies

Even during the Cold War, the United States and the Soviet Union cooperated on nuclear safety and security. Since accidental or unauthorized nuclear detonations anywhere threaten peace everywhere, it seems straightforward that states more experienced in developing nuclear safety and security technologies would transfer such methods to other states. Yet, the historical record is mixed. Why? While existing explanations focus on the political costs and proliferation risks faced by the transferring state, this article argues that specific technological features condition the feasibility of assistance. For more complex nuclear safety and security technologies, robust technical cooperation is crucial to build the necessary trust for scientists to transfer tacit knowledge without divulging sensitive information. Leveraging elite interviews and archival evidence, my theory is supported by four case studies: US sharing of basic nuclear safety and security technologies with the Soviet Union (1961–1963); US withholding of complex nuclear safety and security technologies from China (1990–1999) and Pakistan (1998–2003); and US sharing of complex nuclear safety and security technologies with Russia (1994–2007). My findings suggest the need to examine not only the motivations behind nuclear assistance but also the process by which it occurs and the features of the technologies involved, with implications for how states cooperate to manage the global risks of emerging technologies.

Experimental Study on Two-Phase Countercurrent Flow Limitation in Horizontal Circular Pipes

The two-phase countercurrent flow limitation (CCFL) in horizontal channels is important in relation to nuclear reactor safety. In this study, we aim to investigate the CCFL characteristics and the flow behaviors in horizontal circular pipes with small diameters. The effects of pipe diameter and the water head in the upper plenum on CCFL characteristics are experimentally studied. An image-processing technique and statistical treatments are implemented to analyze the horizontal countercurrent flow. The results show that the CCFL characteristics for the horizontal circular pipes with small diameters can be well correlated using the dimensionless parameters, which are based on adding fluid viscosity to the Wallis parameters. The CCFL characteristics are significantly affected by the pipe diameter and are slightly affected by the water head above the horizontal pipe. The gas–liquid interface fluctuates with certain periods, and flow pattern transitions happen in the horizontal air–water countercurrent flow. As the air flow rate increases, the occurrence location of the liquid slug appears to shift towards the water entrance. In addition, the further away from the water entrance, the lower the average of liquid holdup.

Simulated safety analysis of a total loss of feedwater accident in the steam generator of CPR1000 nuclear power plant

As M310 series units, CPR1000 Nuclear Power Plant in China is also the first batch of demonstration nuclear power plants in China to try out SOP(the state-oriented procedure) accident procedures. The establishment of SOP accident regulations is an important work to mitigate the consequences of radioactive accidents in the design of nuclear power plants, and it is also the embodiment of the defense-in-depth requirements in the implementation of nuclear safety regulations. It can handle superimposed accidents, human errors in program implementation and events not considered in the primary cause sequence. In order to study the efficiency of its upgraded SOP procedure and its ability to deal with severe accidents, the typical primary cause event of severe accidents in nuclear power plants–such as the total loss of feedwater accident of steam generators has been adopted for research. The initial state of the simulation is the full power operation of the nuclear power plant, the simulation process is from the loss of all the water supply accidents to the whole accident processing. A total of 30 simulation experiments were carried out. The temperature and pressure of the key parameters obtained are the average values of 30 simulation experiments. The stability and efficiency of the updated SOP are evaluated by the treatment time of the water supply loss accident and the changes of reactor temperature and pressure after the accident. The results show that the upgraded SOP procedure is more stable and efficient. The upgraded SOP procedure, when dealing with other types of accidents, has improved the treatment of residual heat conduction in the core, increased the efficiency of accident treatment, and enhanced the safety of the reactor. At the same time, based on the weaknesses of the upgraded SOP procedure, constructive suggestions on changing the setting value of the steam generator bypass valve have also been put forward accordingly. Furthermore, it can be concluded from many simulated accidents that this change will win about 5% more of the disposal time for nuclear power plants under accidents, which provides references for the subsequent upgrading of M310 series power plant procedures.

The volcanic aspect on determining Site of nuclear power plant in Indonesia: Gap analysis between standard and regulations

The development of nuclear power plants is in three phases. The first phase is a consideration before the decision on the NPP construction program is approved, the second phase is the preparatory work for making contracts and preparing for the construction of NPP after the NPP construction policy is approved, and the third phase is contracting, licensing and building the first NPP. As a volcanically active country, Indonesia contains over 130 active volcanoes that are part of the Pacific Ring of Fire. The volcanic aspect is one of the safety factors considered while deciding the location of an NPP. Research on the potential of natural external risks to the determination of nuclear power plants in Indonesia, including the volcanic aspect, has been conducted based on the safety reference or safety guide of the IAEA and the Nuclear Energy Regulatory Body (BAPETEN) Regulation. Due to technological advancements, safety needs have evolved so the existing Indonesia National Standard (SNI) must be updated to comply with BAPETEN regulations. The substance in SNI 18-2034-1990 relating to volcanic features seems less relevant in actual conditions, given that more complete and exact criteria for determining a site guarantee the safety and health of residents and surrounding the environment site. The study intends to conduct a gap analysis of volcanic issues in SNI and volcanic regulations. The method used is identification requirements for volcanic aspects in SNI 18-2034-1990 about Determining Site of Nuclear Reactor Guidance with BAPETEN Chairman Regulation (BCR) number 4 of 2018 about Nuclear Installation Site Evaluation Safety Provisions and BCR number 5 of 2015 about Evaluation of Nuclear Installation Sites for Volcanic Aspects, and analysis uses a qualitative method of inductive techniques. The outcome of this research applies to suggesting a revision of SNI number 18-2034-1990, especially the volcanic aspect.