Passive Safety Features Research Articles

The evolution of the Indian nuclear power programme

The strategy for growth of nuclear power in India was planned nearly sixty years ago, noting the rather small uranium and large thorium reserves in the country. This has prompted India to adopt the well-known three stage programme. The first stage is primarily based on the Pressurised Heavy Water Reactors (PHWRs). The evolution of technology of PHWRs is discussed in the first part of this paper. India has constructed eighteen PHWRs, has achieved impressive availability factors and some of these reactors have achieved annual capacity factors of nearly hundred percent in the recent past. Having installed a nuclear power generation capacity of 6780 MWe (4460 MWe from PHWRs, 2000 MWe from PWRs and 320 MWe from BWRs), India is now poised to launch a major expansion programme. This will be based on the increased availability of uranium from import and from the augmented domestic supply. In the immediate future, the nuclear power capacity will grow by installing a series of indigenous PHWRs in addition to light water reactors built under international civil nuclear cooperation agreements. The growth of nuclear capacity in this period is aimed at increasing the share of nuclear power in meeting the base-load demand of non-carbon electricity required for the rapid economic growth in the country.India embarked on its second stage programme with the successful operation of a research reactor named Fast Breeder Test Reactor (FBTR). Based on the experience of FBTR and following the development of all the required enabling technologies in India, the Prototype Fast Breeder Reactor (PFBR) of 500 MWe (gross) capacity has been designed, constructed and is now at an advanced stage of commissioning. A large increase in the nuclear power generation capacity is envisaged through deployment of fast breeder reactors. These reactors will not only help in building up nuclear power capacity but also, in due course, enable conversion of thorium into fissile U233, which will fuel the reactors in the third stage of India's nuclear programme. The adoption of the closed fuel cycle for both thermal and fast reactors has dual objectives: multiplying the fissile inventory by fertile to fissile conversion and reducing the burden of long lived radioactive waste-both being essential for attaining near sustainability of nuclear power.India has designed an Advanced Heavy Water Reactor (AHWR). This reactor can use one of the following fuels:(i)Low enriched uranium dioxide-thorium dioxide based mixed oxide (MOX) fuel primarily in once through mode.(ii)Plutonium dioxide – thorium dioxide MOX fuel in the closed fuel cycle mode. Eventually when enough of U233 is generated the reactor can operate with the U233– thorium MOX fuel.A large number of passive safety features are incorporated in this design.As a part of the long term objective of the development of Accelerator Driven Sub-critical Systems (ADSS), work on the development of high power linear accelerator has been initiated.

Dynamic Response of AP1000 Nuclear Island Due to Safe Shutdown Earthquake Loading

AP1000 is a standard nuclear power plant developed by Westinghouse and its partners by using an advanced passive safety feature. Among the five principle building structures, namely the nuclear island, turbine building, annex building, diesel generator building and radwaste building, the safety of the nuclear island building is the most concerned. This paper investigates the dynamic response of the nuclear island building of the AP1000 plant subjected to safe shutdown earthquake loadings. A finite element model for the building, which is assumed to be built in a hard-rock base, is developed and its dynamic response is computed with the aid of the commercial finite element package ANSYS. The dynamic characteristics, including the natural frequencies, the vibration modes, and the time histories for displacements, velocities, and accelerations of the building are obtained for two typical safe shutdown earthquakes, El Centro and Kobe earthquakes. The dynamic behavior of the building due to the earthquakes and its safety is examined and highlighted.

Experimental Investigation on Characteristics of Boiling Two-Phase-Flow Instability in a Parallel-Multichannel Natural-Circulation System

Natural-circulation systems are gaining wide acceptance due to their inherent passive safety features, which makes them more reliable to use in nuclear power plants. In view of this, many Generation III and III+ nuclear reactors have been designed with natural circulation as a mode of core cooling. But, the major area of concern in these kinds of systems is still the complex phenomenon of boiling two-phase-flow instabilities, which is yet to be fully understood especially when power and pressure in the system are low (type I instability). A major factor in this regard is to know the sensitivity of the number of parallel channels to the characteristic behavior of these flow oscillations in systems like those of a boiling water reactor. Based on mathematical models, in the past, some authors reported that any number of channels behaves in the same way as a twin-channel system. There is no experimental study to validate this. This experimental investigation has been done to add insight. A parallel-multichannel closed loop filled with water and maintained at atmospheric pressure was used for the study. Power in the individual heated sections of the loop was increased from 0 to 2 kW in steps of 250 W. Each power level was maintained for ~30 min. After reaching 2 kW, power was decreased to 0 kW. Three cases of experiments were done by taking two, four, and six active channels at a time, respectively. Different flow oscillation parameters such as amplitude, frequency, phase difference, general characteristics, etc., were studied to see if they were affected when the number of parallel channels was changed. The present analysis showed that their behavior is sensitive toward changes in the number of parallel channels. We cannot extrapolate twin-channel data when there are more channels in a system. The present paper discusses the experiments performed and the detailed results in support of this argument.

Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

Investigation of a “Sharps” incident

Special nuclear material research, process development, technology demonstration, and manufacturing capabilities are provided at the Los Alamos National Laboratory Plutonium Facility. Engineered barriers provide the most effective protection from radioactive and hazardous materials. The Worker Safety Security Team augments these passive safety feature by investigating incidents to identify appropriate prevention and mitigation measures. “Learning Teams” facilitate employee feedback loop and integration toward process improvement. This article reports an investigation of a “Sharps” incident and reviews a case study of a technician that cuts his left thumb while making a gasket. Causal analysis of the sharps incident uncovered contributing factors that created the environment in which the incident occurred. Latent organizational conditions that created error-likely situations or weakened defenses were identified and controlled. Effective improvements that reduce the probability or consequence of similar sharps incidents were implemented.

New reactor cavity cooling system (RCCS) with passive safety features: A comparative methodology between a real RCCS and a scaled-down heat-removal test facility

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features and without electrical or mechanical driving requirement is proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region (RPV region) and a cooling region with a heat-transfer surface to ambient air assumed to be 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV through radiation and natural convection. Employing air as the working fluid and ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay-heat-removal. In the paper, we address a comparative methodology between a real RCCS (REAL) and a scaled-down heat-removal test facility (SCALE). As an example, the heat flux by radiation of (SCALE) can be the same as that of (REAL) because the two have the same temperatures of the RPV and the RCCS, TRPV(REAL)=TRPV(SCALE), TRCCS(REAL)=TRCCS(SCALE), and the same view and configuration factors. Next, we also conduct comparisons on natural convection using the Grashof number, GrrRCCS-rRPV. Here, the ratio of (SCALE) to (REAL) in terms of characteristic length is defined as (rRCCS-rRPV)(SCALE)(rRCCS-rRPV)(REAL)=1x. When the working fluid in (SCALE) can be pressurized up to x32 times that of (REAL), (SCALE)’s Grashof number, GrrRCCS-rRPV, can be the same as that of (REAL). Moreover, (SCALE)’s heat flux on the RPV surface as the experimental conditions for obtaining the same Grashof number as that of (REAL) can be determined. The difference between (SCALE) and (REAL) in terms of the heat flux on the RPV surface is (x−1) times (REAL)’s heat flux by natural convection. Thus, (SCALE) will obtain valuable experimental data for demonstrating the new RCCS’s performance.

Impact of Rare Earth Recovery Fraction on Core Physics Parameters of the Korean SFR for TRU Transmutation

An advanced sodium-cooled fast reactor concept has been developed in Korea for transuranics (TRU) transmutation with an electricity output of 600 MW(electric) (called the KALIMER-600 TRU burner). The core design philosophy is primarily based on passive safety mechanisms to meet the Generation IV technology goals. Accordingly, metal fuel has been adopted to enhance its inherent passive safety characteristics. The charged fuel in a ternary metal alloy (U-TRU-Zr) consists of self-recycled TRU and TRU recovered from the spent nuclear fuels of current light water reactors through a pyro-metallurgical process, which is assumed to carry over 5% of the inventory of rare earth (RE) elements. It has been recognized that an additional amount of RE in the fuel would decrease the material attractiveness of the charged fuel with respect to proliferation resistance and physical protection (PR&PP). However, this may raise concerns because most of the reactor physics parameters will tend to negatively affect the passive safety features encoded in the original core concept. Thus, this study investigates the impact of the RE recovery fraction on the core physics performance and important safety parameters such as Doppler coefficients and sodium void reactivity. The results are expected to help provide guidance regarding the development of limiting conditions for RE contents to recycling technology flow sheet developers and ternary metal fuel developers, and to provide insight into optimizing the core passive safety characteristics under accident conditions should a significant amount of RE be needed to enhance PR&PP.

Experimental simulation of liquid entrainment in ADS-4 depressurization line in AP1000

The fourth stage Automatic Depressurization System is an important passive safety feature in Westinghouse AP1000 which enables controlled depressurization of reactor coolant system in small break LOCA. However, large amount of coolant can be carried to the containment via the ADS-4 branch entrainment and the upper plenum entrainment during the depressurization process, which poses great threats to core uncovering and melting. The automatic Depressurization and Entrainment TEst Loop (ADETEL) modeled after AP1000 with a scaling ratio of 1:5.6 was constructed to investigate the entrainment and depressurization behavior after the actuation of ADS-4 valves. The entrainment and depressurization features were investigated under different initial pressure, mixture liquid level in the pressure vessel and heating power. The entrainment deposition effect of the reactor internals was also investigated. The test data reveals that large amount of water are entrained through the ADS-4 branch line within a short period of time. The liquid entrainment rate and the reduced rate of the mixture liquid level in the pressure vessel increase with the initial system pressure. It is notable that the core uncovery was experienced when the initial pressure was set to 0.5 MPa in current experimental conditions. The reactor internals have little effect on the entrained mass and the mixture liquid levels in the pressure vessel.

Severe External Hazard on Hypothetical JSFR in 2010

Severe external hazards on the Japan Sodium-cooled Fast Reactor (JSFR) have been analyzed and evaluated. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. Integrity of the major components has been confirmed covering recent earthquake conditions. In the case of a tsunami, the seawater pumps for the component cooling water system (CCWS) could be damaged by the tsunami, since they are located at sea level. In the JSFR design with full natural convection decay heat removal systems (DHRSs) and an air-cooling emergency gas turbine, safety-grade components are independent of CCWS, and loss of CCWS does not affect reactor cooling. As a conservative case, hypothetical station blackout (SBO) has been evaluated. In the case of SBO, decay heat is removed by natural convection DHRS, but control of the air cooler (AC) damper is lost after the battery power is out. The analysis has revealed that freezing at one of three ACs could happen due to loss of automatic control of AC dampers. However, the time margin to protected loss of heat sink is evaluated to be >10 days. Manual control of the AC damper is also investigated. Transient analyses show that the AC dampers can be controlled manually adopting a simple operation procedure with sufficient operation time. Decay heat cooling in the case of collapse in all air stacks of AC has been evaluated. The result shows that decay heat could be removed maintaining air paths in two of three ACs by accident management. In conclusion, JSFR in the 2010 design version has enough external hazard toughness mainly thanks to passive safety features and a seismic isolation system.

Quantitative and Qualitative Comparison of Light Water and Advanced Small Modular Reactors

In recent years, several small modular reactor (SMR) designs have been developed. These nuclear power plants (NPPs) not only offer a small power size (less than 300 MWe), a reduced spatial footprint, and modularized compact designs fabricated in factories and transported to the intended sites, but also passive safety features. Some light water (LW)-SMRs have already been granted by Department of Energy: NuScale and mPower. New LW-SMRs are mainly inspired by the early LW-SMRs (such as process-inherent ultimate safety (PIUS), international reactor innovative and secure (IRIS), and safe integral reactor (SIR)). LW-SMRs employ significantly fewer components to decrease costs and increase simplicity of design. However, new physical challenges have appeared with these changes. At the same time, advanced SMR (ADV-SMR) designs (such as PBMR, MHR Antares, Prism, 4S, and Hyperion) are being developed that have improved passive safety and other features. This paper quantitatively and qualitatively compares most of the LW- and ADV-SMRs with respect to reactors, nuclear fuel, containment, reactor coolant systems, refueling, and emergency coolant systems. Economic and financing evaluations are also included in the paper. The detailed comparisons in this paper elucidate that one reactor is not superior to the others analyzed in this study, as each reactor is designed to meet different needs.

Strength assessment of SMART design against anticipated transient without scram

SMART is a small sized integral type pressurized water reactor and its design minimizes the potential for an occurrence of transients and accidents. To achieve these purposes, engineered safety features enhance the accident resistance. A large volume of semi-passive pressurizer can accommodate a wide range of pressure transients. Low core power density increases the thermal margin, and negative fuel and moderator temperature coefficients yield beneficial effects on a core self-stabilization, and limit the reactor power during accidents. Considering the enhanced engineered safety features and design characteristics, anticipated transient without scram scenarios are adopted to evaluate the strength of the SMART design. The peak pressure does not exceed 116% of the nominal pressure and the fuel can maintain its integrity for the transient. The intrinsic SMART design characteristics having passive safety features and a large primary coolant inventory can maintain a stable condition safely although the reactor protection and diverse protection systems are not available for the anticipated transient without scram transients.

Impact of power density profile on passive decay heat removal in prismatic HTGR

Reactor designs with passive safety features have been actively developed ever since the Fukushima Daiichi Nuclear Power Plant accident in 2011. In our previous works, we investigated the design of a small, prismatic HTGR for passive decay heat removal and the relations of parameters for successful decay heat removal. For the power density distribution of the HTGR, the Bessel function was used in the radial direction and the cosine function in the vertical direction, assuming a barrel-type cylindrical core. In practice, the power density distribution is usually flattened by introducing burnable poisons or control rods during operation to prevent non-uniform power production throughout the core. The power density profile might have an effect on the passive safety features of the HTGR. Therefore, the influence of a flat power distribution on the passive safety features of a prismatic HTGR was investigated in the present study. Results showed that flattening the power density profile allowed greater thermal power with the same safety performance without changing the reactor dimensions or calculation conditions. On the other hand, a same-power reactor could be designed with a smaller size without worsening its passive safety features, which is highly advantageous from an economical point of view. Hence, it was confirmed that reactor safety characteristics could be improved by implementing uniform power density in the design.

Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor.

Neutronic and thermo-hydraulic analyses of a small, long-life HTGR for passive decay-heat removal

Since the accident at Fukushima Daiichi Nuclear Power Plant in 2011, design concepts for nuclear reactors have been reconsidered with much greater emphasis placed upon passive systems for decay-heat removal. By considering this issue, the design parameter conditions for high temperature gas-cooled reactors (HTGRs) with passive safety features of decay-heat removal were obtained by residual-heat transfer calculation using equations for fundamental heat transfer mechanisms in our previous works. In the present study, the appropriate size of reactor core for a 100 MWt reactor operating at 1123 K of the initial core temperature was found using the conditions. Consequently, neutronics and thermo-hydraulic analyses for the proposed reactor core were performed and the proper optimizations to control the excess reactivity and flatten the change in power peaking factor during operation were done successfully. By the systematic method to decide the core design which satisfies the condition for passive decay-heat removal, a long-life small HTGR concept whose excess reactivity was small during the operation was shown. The small excess reactivity is a significant advantage from the view point of safety in reactivity accident.

In- and ex-vessel coupled analysis of IVR-ERVC phenomenon for large scale PWR

As a key severe accident management strategy for light water reactors (LWRs), in-vessel retention (IVR) through external reactor vessel cooling (ERVC) has been the focus of relevant studies for decades. However, previous studies only investigated the molten pool configurations considered to be in a final steady state mainly for reactors of such as AP600 and AP1000. Furthermore, most of studies performed in the past dealt with analysis for an isolated IVR-ERVC process, without considering the strong coupling between the internal and external reactor pressure vessel (RPV) conditions. This paper addresses the IVR-ERVC issues from a transient perspective using the severe accident code MELCOR for a large advanced passive power plant: a three-loop, 5000MWt scale pressurized water reactor with passive safety features. The analysis is mainly focused on the severe accident transients including core degradation and relocation, molten pool formation and growth, and heat transfer within a molten pool. Furthermore, internal and external RPV conditions are combined together in the IVR-ERVC analysis. MELCOR calculations for lower head heat flux are then compared with critical heat flux (CHF) to assess the effectiveness of IVR-ERVC. The results suggest that lower head heat flux is below the CHF value. Therefore, the IVR-ERVC strategy for this large PWR is considered to be feasible. It was also found that as the reactor power is raised to large scale PWR, new accident sequences may occur during the severe accident evolution, thus leading to a proposal of a completely new molten pool configuration for future studies.

The Dual Fluid Reactor – A novel concept for a fast nuclear reactor of high efficiency

The Dual Fluid Reactor, DFR, is a novel concept of a fast heterogeneous nuclear reactor. Its key feature is the employment of two separate liquid cycles, one for fuel and one for the coolant. As opposed to other liquid-fuel concepts like the Molten-Salt Fast Reactor (MSFR), both cycles in the DFR can be separately optimized for their respective purpose, leading to advantageous consequences: A very high power density resulting in remarkable cost savings, and a highly negative temperature feedback coefficient, enabling a self-regulation without any control rods or mechanical parts in the core.In the current reference design the fuel liquid is an undiluted actinide trichloride based on isotope-purified Cl-37, circulating at an operating temperature of 1000°C. It can be processed on-line in a small internal processing unit utilizing fractional distillation or electro refining. Medical radioisotopes like Mo-99/Tc-99m are by-products and can be provided right away. In a more advanced design, an actinide metal alloy melt with an appropriately low solidus temperature is also possible which enables a reduction of the core size and allows a further increase in the operating temperature due to its high heat conductivity.For the reference design, pure Lead as coolant is the best choice. It yields a very hard neutron spectrum, fostering a very good neutron economy and therefore making the DFR a preferred thorium breeder but also a very effective waste incinerator and transmuter. With its high coolant temperature the DFR achieves the same ambitions as the Generation IV concept of the very high temperature reactor (VHTR), with all its advantages like electricity production with high efficiency and the synthesis of carbon-free fuels, but with overall production costs competitive with today’s refined oil.The specific combination of the liquids in the very high temperature regime requires structural materials withstanding corrosive attacks. Because of the small size of the reactor core the utilization of these expensive materials would have no significant impact on the overall energy (and also economic) efficiency, measured by the EROI (Energy Return on Investment), which is more than 20 times higher than for a light-water reactor (LWR).The DFR inherits the positive properties of the lead-cooled reactor (LFR) and of the MSFR, especially its outstanding passive safety features.

Feasibility study of a dedicated nuclear desalination system: Low-pressure Inherent heat sink Nuclear Desalination plant (LIND)

Abstract In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal–hydraulic and neutronic design requirements. In a thermal–hydraulic analysis using an analytical method based on the Wooton–Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MWth and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

Sensitive analysis of water levels and air intakes on natural frequency of AP1000 nuclear island building considering FSI effects

AP1000 is an advanced generation III+ nuclear power plant which utilizes a series of passive safety features. The main component of AP1000 is nuclear island which consists of containment vessel, shield building and auxiliary building. Water storage tank and air intake are two of the most important components of nuclear island for cooling down the containment vessel when an accident happens. The fluid–structure interaction effects and the locations of air intakes may affect the vibration characteristics of the nuclear island. In a previous paper, the influence of water level and elevation of air intakes on dynamic characteristics of shield building was studied. A more detailed model of AP1000 nuclear island including auxiliary and internal structures should be built and analyzed to obtain more reliable results due to the complex shape and arrangement of AP1000.The purpose of this paper is to investigate the influence of various water levels and different locations of air intakes on natural frequency of the overall nuclear island of AP1000 by modal analysis considering fluid–structure interaction effects. The results of finite element analysis indicate that the first two natural frequencies decrease as the water levels increase while the 3rd and the 4th frequencies are almost constant, and the elevation of air intake has an insignificant impact on the natural frequencies of the nuclear island building. In addition, the height ratios of h1/H and h2/H also influence the vibration characteristics of nuclear island building. The variation of stiffness as location of air intake is very small compared to the integral rigidity of the overall nuclear island.

A reduced-order primary model for modern PWR primary loop analysis - Part I: RELAP5 benchmarking

A large scale PWR (3400 MWt) with passive safety features is considered and first a transient PWR's pressuriser heat and mass transfers are discussed, and then the primary loop is modelled for two types of LOCAs. We have supposed a major scenario which is described herein so that all emergency cooling systems are lost due to an earthquake. Now, the passive ECCS (including passive ACCs and CMTs) starts under opening of depressurisation valves and gravitation acceleration in the primary loop. We examine a Reduced-order Primary Model (RPM) in this paper, and many physical parameters are determined based on RPM. In the current paper, we have verified the RPM with RELAP5.3.2 for a three inch break. Briefly, this paper contains a reduced primary loop model for estimating the dynamical response of the primary based on fundamental governing equations, and RELAP benchmarking.

Design and technology development for small modular reactors – Safety expectations, prospects and impediments of their deployment

The IAEA has seen a significant increase in participation of Member States and their expertise in the IAEA program for small and medium-sized reactors (SMRs) technology development. There are high-level of interest in SMR development and deployment in technology holders and technology user's countries. The trend in development has been towards design certification of small modular reactors, which are defined as advanced SMRs that produce electric power up to 300 MW(e), designed to be built in factories and shipped to utilities for installation as demand arises. The current driving forces of SMR development are, among others: fulfilling the need for flexible power generation for a wider range of users and applications; replacing the aging fossil-fired units; enhancing safety performance through inherent and passive safety features; offering better economic affordability; suitability for cogeneration and non-electric applications, options for remote areas and hybrid energy systems of nuclear with renewable. The majority of the near-term deployment advanced SMRs is the integrated pressurized water reactors (iPWR). Several countries are pioneering in the development and application of transportable nuclear power plants (TNPP), including floating and marine-based SMRs. The innovative features of advanced SMRs that converge into safer, reliable and affordable plant have been discussed. Multiple module plant distinct concepts of operations, licensing process, legal and regulatory framework are the main issues for the SMRs deployment and are delineated. For newcomer countries, however, one of the challenges is how to reconcile the competitiveness issue between advanced SMR designs and potential energy alternatives including large reactors with proven technology. The paper focuses on some impediments in the deployment of such novel designs that need to be resolved through testing and qualification of components and equipment, Research and Development (R&D), training and international collaboration.