Figure 15.1 shows the SMART development program. It is divided into three phases: technology development, technology verification and commercialization. Since 1997, the Korean government has been supporting the development of SMART technology. During this period, fundamental technologies were developed and the conceptual design was performed. After the conceptual design, the basic design was completed in 2002. Following the technology development phase, a SMART pilot plant design project was launched for comprehensive performance verification. Through the SMART-PPS (pre-project service) phase from 2006 to 2007, the SMART system design was further optimized and a 3-year SMART Technology verification and standard design approval (SMART SDA) program was set. In this program, three years were set for standard design development including experiments and tests for the associated technologies, and a set of licensing application documents were also prepared during the same period. One and a half years for a licensing review was followed after the SDA application. Commercialization of the SMART desalination plant will be introduced just after the SDA phase.

The CAREM project involves technological and engineering solutions, as well as several innovative design features that must be properly demonstrated during the design phase. Also specific codes used for modeling systems related with safety issues to obtain design parameters (primary cooling system, reactor core, fuel design, etc.) must be verified and validated against worldwide benchmarks and/or experimental data to build confidence in their results (Mazzi et al., 2012).

Within the CAREM project, the effort has been focused mainly on the nuclear island (inside containment and safety systems) where several innovative design solutions require developments of the first stage (to assure they comply with functional requirements). This comprises mainly: the reactor core cooling system (RCCS), the reactor core and fuel assembly, and the FSS. An extensive experimental plan has been prepared, including the design and construction of several experimental facilities to fulfill the project’s requirements.

A high-pressure natural circulation rig, CAPCN, was constructed to perform dynamic tests of RCCS. Its purpose was mainly to study the thermal-hydraulic dynamic response of the CAREM primary loop, including all the coupled phenomena that may be described by one-dimensional models. This includes the assessment of the calculation codes on models of the rig, and the extension of validated models to the analysis of the CAREM reactor.

The CAPCN rig resembles CAREM in the primary loop (self-pressurized natural circulation) and the SG (helical once-through), while the secondary loop is designed only to produce adequate boundary conditions. Operational parameters are reproduced for intensive quantities (pressure, temperature, void fraction, heat flux, etc.) and scaled for extensive quantities (flow, heating power, cross-sections, etc.). Height was kept approximately on a 1:1 scale. The heating power may be regulated up to 300 kW. The secondary loop pressure and cold leg temperatures are controlled through valves. The pump regulates the flow. The condenser is of air-cooled type with airflow control. The control of the actuators (heaters, valves, pumps, etc.), data acquisition and operating follow up were carried out from a control room, through a PC-based, multi-node software (flexible enough to define any feedback loop).

Many experiments were performed in order to investigate the thermal-hydraulic response of the system in conditions similar to CAREM operational states. The influence of different parameters like vapor dome volume, hydraulic resistance and dome nitrogen pressure was studied. Perturbations in the thermal power, heat removal and pressure relief were applied. The dynamic responses at low pressure and temperature, and with control feedback loops, were also studied. It was observed that around the operating point self-pressurized natural circulation was very stable, even with important deviations on the relevant parameters. A representative group of transients were selected, in order to check computer models.

The thermal-hydraulic design of CAREM reactor core was carried out using a 3-D two fluid model code. In order to take into account the strong coupling of the thermal-hydraulic and neutronic of the core, this code was linked with neutronic codes. This coupled model allows a 3-D evaluation of power and thermal-hydraulic parameters at any stage of the burnup cycle.

The mass flow rate in the core of the CAREM reactor is rather low compared with typical light-water reactors and therefore correlations or experimental data available should be assessed in the range of interest. In order to perform this assessment experiments were conducted at the thermal-hydraulic laboratories of the Institute of Physics and Power Engineering (IPPE, Obninsk, Russian Federation).

The main goal of the experimental program was to generate a substantial database to assess the prediction methodology for critical heat flux (CHF) applicable to the CAREM core, covering a wide range of thermal-hydraulic parameters around the point of normal operation. Most tests were performed using a low-pressure Freon rig, and results were extrapolated to water conditions through scaling models. Finally a reduced set of tests were performed in water at high pressure and temperature, to validate the method for scaling. Different test sections were assembled to simulate different regions in the fuel element as well as radial uniform and non-uniform power generations. A bundle with 35% of the full length was tested to obtain CHF data under average sub-cooled conditions. More than 250 experimental points under different conditions were obtained in the Freon loop and more than 25 point in the water loop.

The fuel assemblies and absorbing clusters were or will be subject to a series of tests, including standard mechanical evaluations, and hydraulic tests. The latter comprise:

• tests in a low-pressure rig evaluating pressure losses (performed), flow-induced vibrations

and general assembling behavior;

• endurance tests in a high-pressure loop points to wear-out and fretting issues.

A fuel rod irradiation test to be performed in the Halden boiling water reactor is under preparation. This test has the purpose of characterizing the most relevant performance aspects of the fuel, such as temperature behavior, dimensional stability and fission gas release.

Neutronic modeling validations were made against VVER reactor geometry using experimental data from a ZR-6 Research Reactor, Hungary, and a series of benchmark data for typical PWR reactors. The RA-8 critical facility has been designed and constructed as an experimental facility to measure neutronic parameters of CAREM. Further experimental data were obtained from RA-8.

One of the most innovative systems behind the CAREM concept is the hydraulic (in-vessel) control rod drive (HCRD) mechanism. The design embraces mechanical and thermal-hydraulic innovative solutions, so a complete experimental program is necessary to achieve the high reliability performance jointly with low maintenance. This development plan includes the construction of several experimental facilities.

Preliminary tests were first performed to prove the feasibility of the theoretical approach, to have an idea of some of the most sensitive controlling parameters and to determine spot points to be focused on during design. Tests were undertaken on a rough device with promising experimental results, and good agreement with first modeling data was obtained.

First prototype tests helped to determine preliminary operating parameters on a full-scale mechanism as a first approach towards detail engineering. These parameters include range of flow, ways to produce hydraulic pulses, etc. Manufacturing hints that simplified and reduce costs of the first design were also found. Tests were carried out in a specially built rig and as part of this experimental development it was decided to separate the regulating and fast-drop requirements in different devices.

Test on a low-pressure loop were carried out with the CRD at atmospheric pressure, and with feed-water temperature regulation up to low sub-cooling. The feed-water pipeline simulated alternative configurations of the piping layout with a second injection line (dummy) to test possible interference of pulses. The ad-hoc test loop (CEM) was designed to allow automatic control of flow, pressure and temperature, and its instrumentation produces information of operating parameters including pulse shape and timing. The tests included the characterization of the mechanism and the driving water circuit at different operating conditions, and the study of abnormal situations as increase in drag forces, pump failure, loss of control on water flow or temperature, saturated water injection, suspended particle influence, and pressure ‘noise’ in feeding line. The tests, carried out at turbulent regime, which are the closest conditions to operation obtained in this loop, showed good reliability and repetitiveness as well as sensitivity margins for the relevant variables within control capabilities of a standard system.

Finally, a high-pressure loop (CAPEM) was constructed in order to reach the actual operating conditions (P = 12.25 MPa, T ^ 326 °C). The main objectives are to verify the behavior of the mechanisms, to tune up the final controlling parameter values and to perform endurance tests. After this stage, the system under abnormal conditions, such as the behavior during RPV depressurization, simulated breakage of feeding pipes, etc., will be tested.

Since the HCRD design adopted has no movable parts outside the RPV, it was necessary to design a special probe to measure the rod position able to withstand primary environmental conditions. The proposed design consists in a coil wired around the HCRD cylinder with an external associated circuit that measure electric reluctance variations induced by the movement of the piston-shaft (made of magnetic steel) inside the cylinder. A cold test performed showed that the system is capable of sensing one-step movement of the regulating CRD, with an acceptable accuracy. In-furnace high-temperature tests were conducted to evaluate the behavior of the system against temperature changes similar to those occurring during operational transients.

To fulfill the Argentinian regulation, CAREM 25 has two independent shutdown systems. Two independent reactor protection systems exist to actuate these shutdown systems and the safety systems. The qualification of these reactor protection systems includes the development, construction and testing of prototypes.

In a current large PWR, the hot RCS water leaving the top of the reactor fuel is directed into the reactor coolant system hot-leg piping. The hot-leg piping, nominally 29-inch piping (74 cm) (NRC, 2006) connects to a U-tube or a once-through steam generator, where heat is transferred from the primary water to the secondary water to generate steam to drive the turbine generator. Subsequently, colder primary water exits the steam generator into the intermediate leg (cross-over leg) piping, nominally 31-inch piping (79, cm) (NRC, 2006) which connects to the suction side of a reactor coolant pump. The intermediate leg level is typically the largest diameter pipe in the RCS and it is below the level of the top of the fuel assemblies, which can make a leak in this section of piping the most problematic of all large-break LOCAs. The reactor coolant pump discharges into the 27.5 inch (70 cm) (NRC, 2006) RCS cold leg which directs primary water back into the reactor pressure vessel. Current large PWRs employ two, three, or four reactor coolant loops. A typical two-loop large PWR reactor coolant system is shown in Figure 5.3.

Since the pressurizer and steam generator functions are integrated into the iPWR pressure vessel, all the large diameter piping associated with current large PWRs is eliminated. This eliminates any possibility of a large-break LOCA in all the iPWR designs. As a result, active emergency equipment, such as high pressure injection pumps, associated with current large PWRs to mitigate the consequences of a large — break LOCA are also eliminated in the iPWR designs.

The small scale and unique environments of iPWRs open up the instrumentation field to new and innovative solutions. The temptation will be to fall back on what has been used before, and in scenarios where this works, perhaps this will be the path of least resistance. Because of design peculiarities and non-standard physical environments, new instrumentation or redesigned instrumentation will likely be required. In addition to qualifying these new designs and new approaches for new measurement technologies, the iPWR design engineer will need to coordinate with the regulator to accommodate the regulator’s concerns/requirements at the same time.

The sections below identify some of the known challenges and trends in the I&C world for the iPWR designer.

In older control rooms, discrete control input devices (that is, devices that depend upon mechanical motion) are still the most common means for the operator to interact with the plant’s systems. These devices are limited to relatively primitive devices like buttons, switches and levers. As I&C and HSI technology evolves it is becoming possible to control plant components by means of direct manipulation devices like a computer mouse, joystick, keyboard or trackball. As computing power continues to increase, we can expect to see even more sophisticated devices finding their way into the control room and other work areas. In the not too distant future we can expect to see fixed as well as mobile devices that allow not only direct interaction through touch and force feedback, but also indirect interaction through gesture, speech and gaze.

• Probability of adversary success — The probability that an adversary will successfully complete a pathway and generate a consequence.

• Consequences — The effects resulting from the successful completion of the adversary’s intended action described by a pathway, including the effects of mitigation measures.

• Physical protection resources — The staffing, capabilities, and costs required to provide PP, such as background screening, detection, interruption, and neutralization, and the sensitivity of these resources to changes in the threat sophistication and capability.

Measures can be estimated with qualitative and quantitative approaches, which can include documented engineering judgment and formal expert elicitation. Measures can also be estimated using probabilistic methods (such as Markov chains and event trees) and two-sided simulation methods (such as war-gaming techniques). GenlV International Forum (2011b) reviews a number of system analysis techniques relevant for PR&PP studies.

SMRs are defined broadly to include a range of technologies, particularly a range of diverse fuels and coolants. US SMR reactor technology currently chosen for NRC licensing and near-term commercial deployment (2022-2025) is based on LWR technology. Most of these SMRs will be integral pressurized water reactors (iPWRs) as discussed in Part II of this Handbook. LWRs have a well-established framework of regulatory requirements, a technical basis for these requirements, and supporting regulatory guidance that provides acceptable approaches for meeting NRC requirements. The NRC uses a standard review plan (SRP), NUREG-0800, to review licensing applications for these reactor designs. NUREG-0800 was revised in January 2014 to provide general review guidance for SMRs. The NRC will require design-specific SRPs for the licensing of SMRs. These SRPs are under development by the NRC and SMR vendors. The SRP for the B&W mPower design is expected to be published in mid-2014 to support the submittal of the mPower design certification application scheduled in 2015. Additionally, the NRC has a well-established set of validated analytical codes and methods and a well-established infrastructure for conducting safety research needed to support its independent safety review of an LWR plant design and the technical adequacy of a licensing application. It should be emphasized that the near-term SMRs, particularly the iPWRs, will be subject to the same licensing process and the same safety requirements and standards as new large LWRs resulting in no diminution of safety.

New SMRs can be licensed under either of two existing regulatory approaches. The first approach is the traditional ‘two-step’ process described in Title 10, Part 50, ‘Domestic Licensing of Production and Utilization Facilities,’ of the Code of Federal Regulations (10 CFR Part 50), which requires first a construction permit (CP) and then a separate operating license (OL). The second approach is the new ‘one — step’ licensing process described in 10 CFR Part 52, ‘Licenses, Certifications, and Approvals for Nuclear Power Plants,’ which incorporates a combined construction and operating license (COL). It should be noted that the both the SMR vendor B&W and the proposed license applicant TVA have chosen the two-step Part 50 process for licensing of the mPower SMR at the Clinch River, Tennessee site. This two — step process was chosen by TVA and approved by the NRC for this first-of-a-kind (FOAK) design to reduce the risk of delay in obtaining a final design certification under the Part 52 process. It is assumed that all other U. S. applicants for FOAK SMRs and nth of a kind SMRs will adhere to the Part 52 NRC licensing process.

Key to the licensing of any new reactor design, including SMRs, is the approach that the regulatory authority will utilize in assessing the safety basis for the reactor and associated safety systems. Regardless of the licensing process chosen, the licensing authority will need to consider the relative merits of a deterministic versus a risk-informed performance-based approach to assess and approve the safety case for SMRs.

The deployment options for a small reactor build on one of the key attributes of their design. The small modular reactor (SMR) factory build aspect creates a number of significant advantages during the deployment phase. Delivery of assured modules to a customer site has been claimed as a banner headline, but it is worth reviewing the aspects behind these individual claims to understand the contribution of their worth to the overall deployment:

• time to build;

• risk — schedule risk and cost over-run;

• validation — off-site pre-delivery testing;

• international aspirations;

• Critical path through site activities with modular build.

12.5.1.1 Modularity: addressing schedule and cost risk

For the large part nuclear construction project cost increases and schedule delays have their root in site uncertainty when compared to original plans and cost estimates. Modularity has been offered as the solution to these over-runs.

The Construction Industry Institute defines modularity as ‘work that represents substantial offsite construction and assembly of components and areas of the finished product.’ There are incremental levels of modularity that need to be considered; Graduations of modular construction are as follows:

• Skid — modular skid assemblies of equipment that are delivered to site fully instrumented and validated, requiring little installation and commissioning activity. Within this definition it would also be anticipated that the skids are capable of being delivered to site using conventional transportation networks.

• Super skid — this represents the next increment with larger portions of plant intended to be installed in a building.

• Prefabricated — this is the final step in the definition of modular and would extend to a fully prefabricated module comprising plant equipment in an architecturally complete housing.

Has a modular build approach been deployed already with large nuclear plants? There are elements of modular construction encompassed within large nuclear plants. The Vogtle plant built by Southern Co. in Georgia has utilised a level of off-site modular manufacture. This facility has been established to build modules for the plant in a manner that has the same overall goal. However, the modules for a large nuclear plant are essentially one-off modules, built off-site in a piece-by-piece approach. Without the repeat volume flowing through such a facility it can be argued that module build for a large nuclear plant is an off-site variation to the on-site stick build approach and falls short on delivering some of the goals for the module-build approach advocated for small reactors.

The incorporation of advanced modular construction centres around completely shop-fabricated modules. This method of manufacture is more demanding in terms of upfront planning and management. With greater emphasis on design integration, procurement, manufacturing and construction activities there is less natural opportunity for schedule variation. This modular method of site deployment marries beautifully with the flowline approach bringing about greater schedule adherence.

This modular deployment model is not without consequence. The preliminary and final design activities must remain sealed to allow time for the procurement and delivery activities to be completed. Design change at this point erodes the benefits of the advanced modular approach. The mindset of deployable small reactor manufacturing is driving the product line closer and closer towards commodity-based supply chain design.

Having taken a view on the supply and delivery end of the deployment model, there are equivalent considerations that need to be taken into account with the site implementation.

Preliminary, generalized analyses of proposed hybrid system architectures have been conducted by researchers at the INL to provide initial insight into the value of system hybridization (see refs [3], [4], [18-27]). More detailed analyses of tightly coupled HES requires detailed subsystem models accurately to predict dynamic system performance and to identify potential operating procedures to best manage and mitigate variable energy generation sources and to alternate between output products (e. g. electricity and a chemical process). As the tools for dynamic hybrid systems analysis become more sophisticated, and conceptual subsystems and integrated system designs are better defined, more accurate dynamic analyses can be performed to determine optimum system configuration and specific operating parameters, to design integrated system control architecture, and to establish instrumentation requirements for state estimation necessary for control.

13.2 Future trends

Expanded use of nuclear energy for industrial applications raises several technical challenges that can be addressed through conceptual design; system modeling and optimization; component, subsystem, and integrated system testing (non-nuclear and nuclear); and eventual deployment of the advanced energy system. Early work should be focused on identifying key challenges and then addressing those challenges via design solutions. Advanced modeling and simulation tools can provide significant insight to system feasibility and anticipated performance, beginning with simplified steady-state analysis and moving to more complex dynamic analysis with more detailed subsystem models. The more detailed, dynamic system model can then be used as a virtual test bed for control system design prior to hardware implementation. However, many of the significant challenges associated with hybrid systems derive from complex integrated system control and hardware interfaces, including fast­switching smart valves necessary to divert thermal energy between output streams and connections to energy storage systems. These challenges can be addressed through a number of component and subsystem tests for model validation, followed by integrated system testing.

The SMART core is composed of 57 fuel assemblies, the design and performance of which are based on the proven 17 X 17 array with UO2 ceramic fuel rods in the commercial PWRs. Each fuel assembly holds 264 fuel rods of 2 m in active height, 24 guide tubes for control rods, and one instrumentation thimble, which are mechanically joined in a square array. A total of eight space grids hold the fuel rods in their specified positions. A specially designed bottom end piece offers an improved resistance to the debris entering the core. Slightly enriched uranium oxide of less than 5 wt% is used as the fuel ingredient, providing a sufficient amount of reactivity required for a 36 month cycle operation. The fuel assembly is designed to accommodate power ramps during the load-following maneuvering normally experienced in a commercial PWR. The maximum burn-up of the SMART fuel rod is estimated as 54 000 MWD/ MTU. In order to reduce the critical boron concentration and to control the radial power peaking, a lumped burnable absorber rod comprised of Gadolinia (Gd2O3) mixed in Urania (UO2) of which the 235-U enrichment is 1.80 wt% is used. Eight to 24 burnable absorber rods are used per fuel assembly.