The SPES3 Facility for Testing an Integral Layout SMR: BDBE Simulation Analysis

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.

Conceptual Design of Small-Sized HTGR System for Steam Supply and Electricity Generation (HTR50S)

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a small-sized HTGR for steam supply and electricity generation (HTR50S) to deploy the high temperature gas cooled reactor (HTGR) in developing countries at an early date (i.e., in the 2030s). Its reactor power is 50MWt and the reactor outlet temperature is 750°C. It is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and the district heating, and electricity generation using a steam turbine. The design philosophy of the HTR50S is to upgrade the performance from the Japanese first HTGR (HTTR) and to reduce the cost for the commercialization by utilizing the knowledge obtained by the HTTR operation and the design of an advanced commercial plant of 600 MWt-class Very High Temperature Reactor (GTHTR300 series). The major specifications of the HTR50S were determined based on its design philosophy. And the targets of the technology demonstration using the HTR50S for the future commercial small-sized HTGR were identified. The system design of HTR50S was performed to offer the capability of electricity generation, cogeneration of electricity and steam for a district heating and industries. The market potential for the small-sized HTGR in the developing countries was evaluated for the application of the electricity, process heat, district heating and pure water production. It was confirmed that there is enough market potential for the small-sized HTGR in the developing countries. This paper described the major specification and system design of the HTR50S and the market potential for the small-sized HTGR in the developing countries.

Role of Residual Stresses and Strains Fields Generated by Heat Treatments on the Hydrogen Embrittlement of a Nuclear Reactor Pressure Vessel

The wall of a nuclear reactor pressure vessel can undergo a reduction of its mechanical properties due to the presence of hydrogen, a process known as hydrogen embrittlement (HE). A numerical model of hydrogen diffusion assisted by stress and strain was used in this paper to evaluate the HE process in the wall of a real nuclear reactor pressure vessel, formed by a bimaterial (stainless steel and low carbon steel). In this sense, a quantitative analysis was carried out of the influence of tempering heat treatments conditions applied to these two steels on hydrogen concentration accumulated in the nuclear reactor vessel during its operation time. To this end, the most relevant parameters of these heat treatments were considered: (i) tempering temperature and (ii) tempering time.

Combined CFD and DEM Simulations of Fluid Flow and Heat Transfer in a Pebble Bed Reactor

This paper presents combined computational fluid dynamics (CFD) and discrete element method (DEM) simulations of fluid flow and relevant heat transfer in the pebble bed reactor core. In the pebble bed reactor core, the coolant passes highly complicated flow channels, which are formed by thousands of pebbles in a random way. The random packing structure of pebbles is crucial to CFD simulations results. The realistic packing structure in an entire pebble bed reactor (PBR) is generated by discrete element method (DEM). While in CFD calculations, selection of the turbulence models have great importance in accuracy and capturing the details of the flow features, in our numerical simulations both large eddy simulation (LES) and Reynolds-averaged Navier-Stokes (RANS) models are employed to investigate the effects of different turbulence models on gas flow field and relevant heat transfer. The calculations indicate the complex flow structure within the voids between the pebbles.

FCM Fuel Development for SMR Applications

Fully ceramic microencapsulated (FCM) fuels are identified as suitable fuel forms for development and deployment of SMR technologies. High thermal conductivity of the composite fuel coupled with its exceptional radiation and environmental stability enable monolithic compact cores for SMR applications. Meanwhile the built-in redundancy of multiple layers for fission-product retention alleviates the need for structurally demanding pressure vessels and containment buildings. All of the above will enable the reactor designer to achieve a compact and passively safe SMR through utilization of a flexible fuel technology that is currently under active development.

Hydroelastic Modal Analysis of the Perforated Cylindrical Shell in SMART

The System-integrated Modular Advanced ReacTor (SMART) developed by KAERI includes components like a core, steam generators, coolant pumps, and a pressurizer inside the reactor vessel. Though the integrated structure improves the safety of the reactor, it can be excited by an earthquake and pump pulsations. It is important to identify dynamic characteristics of the reactor internals considering fluid-structure interaction caused by inner coolant for preventing damage from the excitations. Thus, the finite element model is constructed to identify dynamic characteristics and natural frequencies and mode shapes are extracted from this finite element model.

Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

Adaptive Identification and Control for Small Modular Reactors

This paper discusses adaptive identification and control (AID&C) techniques to enable automated online identification and control of SMRs. Adaptive system ID allows engineers to rapidly measure system dynamics, calibrate sensors channels, determine loop processes, and quantify actuator authority for the various reactor control loops. Adaptive control can automatically tune these loops and adjust plant processes to optimize conditions for peak performance and power production. Another advantage of the adaptive ID and control approach is that these tools can be used during reactor operation to monitor active and passive components. Adaptive system ID techniques are used to measure loop-transfer characteristics. Presented is a practical approach that uses adaptive model-matching tools to identify the coprime factors of the local loops. This has the advantage over model based approaches since coprime factors can be identified on the real system using real data. Adaptive control enables auto-tuning of controller parameters to meet operational specifications. Using the coprime factors, all controllers that stabilize the plant can be parametrized by a free Q-parameter that can be changed to meet control system objectives and improve performance, and the tuning is performed using adaptive techniques. The controller architecture presented provides several desirable and necessary features: e.g., a default fail-safe mode of operation, stability in the presence of communications failures, guaranteed stability, and robustness. An advantage of the adaptive structure presented here is that control system stability can be guaranteed, even during adaptation by ensuring certain norm conditions on the Q-parameter and estimated plant uncertainty. More importantly, the Q-parameter can be monitored during operation, providing a real-time estimate of the changes in the plant resulting from changes in the reactor itself. This signal monitors the dynamics of each loop, providing information about the reactor from the perspective of the control process. Online monitoring using AID&C can be used to better track control system transients that result in reactor trip, thus avoiding undesirable reactor trips and diversion events. And, there is a potential that the system can better adapt to changing operating conditions during plant transients including load following procedures.

Digital Voltage Regulators and Speed Controllers for Class 1E Safety Related Applications

This paper discusses the design and implementation of both a Voltage Regulator and Speed Controller that are designed to meet the stringent requirements of a Nuclear Power Plant environment. These control systems will be utilized along with an Emergency Diesel Generator (EDG) to enable safe and seamless operation of the power plant in the event of a power grid failure. Although Small Modular Reactor (SMR) plants operate at lower power levels and require less cooling after shutdown, there remains a need for an Emergency Diesel Generator to deal with various accident scenarios and/or Loss of Offsite Power (including extended outages). During normal operation the power plant operational control systems will be powered by the power grid. In the event of a grid failure the EDG, which is controlled by the Voltage Regulator and Speed Controller, will be brought on line to supply power to the power plant’s operational control systems. This paper describes the design requirements and the key features of the Voltage Regulator and Speed Controller design that allow it to function in a “safety” critical application. The overall system is a Class 1E rated system (including Class 1E seismic requirements) and is rated to operate continuously over a 0–50°C ambient air temperature range.

Seismic Protection of Small Modular Reactors

The next edition of ASCE Standard 4 will include detailed provisions for the seismic isolation of structures, systems and components in safety-related nuclear structures. The provisions are based on those available in North America for buildings, bridges and other infrastructure but address issues particular to nuclear energy construction and take advantage of recent research funded by federal agencies, including the Nuclear Regulatory Commission and the National Science Foundation. The paper highlights these research products and their incorporation into ASCE Standard 4. Although the focus of the studies and ASCE Standard 4 is analysis of conventional light water reactors of 500+MWe, most of the conclusions are applicable to small modular reactors.