Status of the small modular fluidized bed light water nuclear reactor concept

The state of the art of a small modular reactor concept with a suspended core is presented. The reactor design is based on a fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident conditions. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another option is to modify the shape of the reactor core to produce a non-fluctuating bed and, consequently, guarantee the dynamic stability of the reactor. The mixing of tantalum in the fuel is also proposed as an additional inhibition to the power excursion. The spent fuel pellets may not be considered nuclear waste, since they are of a shape and size that can easily be used as a source of radiation for food irradiation and industrial applications. The reactor can easily operate as a plutonium burner or can operate with a thorium fuel cycle.     

Cermet fuel in a light water reactor: a possible way to improve safety. Part II. Testing the fuel

This paper summarizes what is done for the experimental testing of cermet fuel with various matrix materials. Low neutron absorption, high heat conductivity, good corrosion resistance in water, low chemical interaction with cladding (zirconium alloy) and UO₂ in normal and accident conditions, technological ability — are the requirements of the matrix material [1]. Suitability of the proposed solutions to the cermet fuel design with respect to these requirements was proven through a series of experiments simulating fuel operating and accidental conditions.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Addressing the challenges posed by advanced reactor passive safety system performance assessment

Passive safety features play an essential role in the development of nuclear technology and within advanced water cooled reactor designs. The assessment of the reliability of such systems in the frame of plant safety and risk studies is still an open issue. This complexity stems from a variety of open points coming out from the efforts conducted so far to address the topic and concern, for instance, the amount of uncertainties affecting the system performance evaluation, including the uncertainties related to the thermal-hydraulic (T-H) codes, as well as the integration within an accident sequence in combination with active systems and human actions. These concerns should be addressed and conveniently worked out, since it is the major goal of the international community (e.g. IAEA) to strive to harmonize the different proposed approaches and to reach a common consensus, in order to add credit to the underlying models and the eventual out coming reliability figures. The main key points that may influence the reliability analysis are presented and discussed and a viable path towards the implementation of the research efforts is delineated, with focus on T-H passive systems.     

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR⁺ (ASBWR⁺) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Long-term aging of light water reactor concrete containments

This paper evaluates the aging of light water reactor concrete containments and identifies three degradation mechanisms that have the potential to cause widespread aging damage after years of satisfactory experience: alkali–silica reactions; corrosion of reinforcing steel, steel liner, and prestressing steel; and sulfate attack. The aging evaluation is based on a comprehensive review of the relevant technical literature. Low-alkali cement and slow-reacting aggregates selected according to ASTM requirements cause deleterious alkali–silica reactions. Low concentrations of chloride ions can initiate corrosion of the reinforcing steel if the hydroxyl ions are sufficiently reduced by carbonation, leaching or magnesium sulfate attack. Magnesium sulfate attack on concrete can also cause loss of strength and degradation of cementitious properties of the containment concrete after long-term exposure. The techniques for inspecting, mitigating and repairing these long-term aging effects are discussed.     

Two new passive safety systems for LWR applications

In a research activity that SIET has been conducting for years about safety systems for light water reactors (LWRs), attention has been paid to developing two passive injection systems representing an innovative solution in mitigating the consequences of loss of coolant accidents. Both systems allow the completely passive injection of cold water into a pressurised vessel. They are triggered by a low-level signal and work on the base of phenomena like natural circulation and condensation. The simplest system, Sistema Iniezione Passiva 1 (SIP-1), injects water contained in a tank into a circuit at the same pressure as the circuit. The most complex system, injection cyclic system (ICS), injects cold water, by filling cyclically a proper tank with the water stored in an atmospheric pressure pool. Thanks to the ENEA sponsorship, this activity has been conducted in three steps: the definition of the conceptual design of the systems; the application of the Relap5 code to simulate their behaviour; and the proposal of their specific applications to pressurised and boiling LWR. In this paper, both systems are presented in their structural and operating characteristics together with the main results of the code application for their simulation. Some proposals of application of SIP-1 and ICS to pressurised water reactors and boiling water reactors are also shown. The developments and reached goals of the prosecution of the research are also summarised here, together with future needs.     

PbBi-cooled direct contact boiling water small reactor

A design concept of PbBi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot PbBi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for PbBi circulation. The generated steam passes through separators and dryers for the removal of PbBi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a PbBi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of PbBi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5%dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with PbBi. As a result, the PbBi core outlet temperature becomes 460°C. The PbBi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.     

Assembly homogenization techniques for light water reactor analysis

Recent progress in development and application of advanced assembly homogenization methods for light water reactor analysis is reviewed. Practical difficulties arising from conventional flux-weighting approximations are discussed and numerical examples given. The mathematical foundations for homogenization methods are outlined. Two methods, Equivalence Theory and Generalized Equivalence Theory which are theoretically capable of eliminating homogenization error are reviewed. Practical means of obtaining approximate homogenized parameters are presented and numerical examples are used to contrast the two methods.

Applications of these techniques to PWR baffle/reflector homogenization and BWR bundle homogenization are discussed. Nodal solutions to realistic reactor problems are compared to fine-mesh PDQ calculations, and the accuracy of the advanced homogenization methods is established. Remaining problem areas are investigated, and directions for future research are suggested.     

Radiation-induced embrittlement in light water reactor pressure vessels

In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integrity is a significant economic consideration because the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant. In addition to plant life considerations, LWR-PV embrittlement creates significant cycle-to-cycle impact through the restriction of normal heat-up and cool-down reactor operations.Recent LWR-PV benchmark experiments are analyzed. On this basis, it is established that an exponential representation accurately describes the spatial dependence of neutron exposure in LWR-PV. Implications produced by this simple exponential behavior are explained and trend-curve models for the prediction of PV embrittlement are derived. These derivations provide for a clearer understanding and assessment of the assumptions underlying these trend-curve models. It is demonstrated that LWR-PV embrittlement possesses significant material dependence.     

Testing the dynamics of shutdown systems instrumentation in reactor trip measurements

Periodic testing of the dynamics of the shutdown systems and their instrumentation is performed in the CANDU nuclear power plants of Ontario Power Generation (OPG) and Bruce Power. Measurements of in-core flux detector (ICFD) and ion chamber (I/C) signals responding to the insertion of shut-off rods (shutdown system No. 1, SDS1), or to the injection of neutron absorbing poison (shutdown system No.2, SDS2) are regularly carried out at the beginning of planned outages. A reactor trip is manually initiated at high power and the trip response signals of ICFDs and I/Cs are recorded by multi-channel high-speed high-resolution data acquisition systems set up temporarily at various locations in the station. The sampling of the seaprate data acquisition systems are synchronized through the headset communication systems of the station. A total of 120 station signals can be sampled simultaneously up to 2500 samples per second. The effective prompt fractions of the ICFDs are estimated from the measured trip response. Effectiveness and the timeline of the trip mechanism are assessed in the measurement as well. The measurement can identify ICFDs with abnormally slow response (under-prompt) or overshooting response (over-prompt) at the beginning of the outage. The time required for the signals to drop to predefined fractions of their pre-trip values (level crossing time) is plotted as a function of detector position and compared against safety requirements. The propagating effect of shut-off rod insertion or poison injection on the flux is monitored by the level crossing times of ICFDs and ion chambers.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

压水堆核电厂实时传输重要安全参数的规范化初步研究

核电厂实时安全参数传输和交换对于核电厂日常管理和应急响应都有重要的意义.基于相关法规标准要求和应急响应需要,结合各机组类型的具体参数,并考虑参数的可获得性,提出了相对统一的PWR核电厂实时传输安全重要机组参数和实时环境监测参数,为核电厂安全相关参数传输标准化工作提供了一定的技术基础.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Integrated modular water reactor (IMR) design

Integrated modular water reactor (IMR) has been developed as one of the advanced small-scale light water reactors, with a thermal output of 1000 MW. The IMR adopts natural circulation and self-pressurization in the primary cooling system, and a reactor vessel built-in steam generators. The core design has been performed using the current light water reactor technology. Thermal-hydraulic sensitivity analyses have been done from the viewpoint of the departure from nucleate boiling (DNB) limitation. The IMR core, with 97 21×21-type fuel assemblies and natural circulation in the primary coolant system, shows a good nuclear and thermal-hydraulic behavior and good allowable margins for the DNB phenomenon. The reactivity change with burnup is about 1%Δk by using burnable absorbers, and only 12 rod-cluster-controls are used through the operating cycle. The 20 m-height reactor vessel encloses steam generators in vapor and liquid portions. Plant dynamic analyses have been also performed in order to evaluate the IMR behavior from the viewpoints of plant operation and control. This study shows that the IMR will operate with enough margins for the core safety and will be stably controlled for load demand changes expected during normal operations.     

Improved core design of the high temperature supercritical-pressure light water reactor

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C.In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained.In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.     

Investigation on pre-oxidation of steels in light water reactor circuits

The state of the art of procedures for the generation of high grade corrosion-protective layers is presented. Quality criteria are discussed. The draft concerning the investigation of pre-oxidation of steels by hot water containing dissolved oxidizing agents is pointed out and substantiated. The evaluation of oxide layers generated in that way is difficult and can be carried out only by using several methods in combination. The validity of gravimetric and surface analytical methods is discussed. Preliminary results of experiments on pre-oxidation by hot water containing oxygen are presented and evaluated.