Small Pressurized Water Reactor Research Articles

Coupling of an innovative small PWR and advanced sodium-cooled fast reactor for incineration of TRU from once-through PWRs

In the nuclear industry, safely managing spent fuels discharged from PWRs (pressurized water reactors) is an ongoing challenge. In this paper, a synergistic coupling of innovative small long-cycle PWRs and advanced sodium-cooled fast reactors is considered to reduce the accumulated TRUs (transuranics) by transmuting them with electricity production. In the coupling strategy, the innovative small PWRs employing UO2–ThO2 and fully ceramic micro-encapsulated fuels are used to deeply burn TRUs from commercial PWRs, while advanced SFRs (sodium-cooled fast reactors) with actinide recycling are designed to further transmute the TRUs discharged from innovative small PWRs. This work focuses on the core physics analysis of new SFR burner cores using different TRU feeds discharged from small PWRs. Additionally, quasi-static reactivity balance analyses are performed to understand the safety of the SFR burner cores. The mass flows of TRUs in the nuclear park, which is composed of PWRs, small long-cycle PWRs, and SFR burners, are analyzed to evaluate TRU inventory reduction. The results of this study show that the advanced SFR burners with all the TRU feed types discharged from the small long-cycle PWRs have a high TRU consumption rate. They satisfy all of the conditions for self-controllability under unprotected accidents with a reasonable number of control rods. This coupling strategy requires ~35% less power in conjunction with the advanced SFR burners in the nuclear park and increases the support ratio of SFR burners by ~42% than does the coupling of commercial PWRs and SFR burners. Copyright © 2015 John Wiley & Sons, Ltd.

Combination of Thermochemical Energy Storage and Small Pressurized Water Reactor for Cogeneration System

Abstract In recent decades, small nuclear reactors for cogeneration systems have been studied. Generally, nuclear plants are operated at steady state, but the heat demand load is not steady. Load leveling using thermochemical energy storage (TcES) is a potential method for overcoming this problem. In this work, the heat load of a town was assumed, and the required amount of TcES material for load leveling was estimated. The amount of material required, relative to the volume of the nuclear reactor vessel, was acceptable for installing a TcES system at the nuclear plant.

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO2–ThO2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO2–ThO2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO2–UO2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods.

Feasibility Study of a Soluble Boron–Free Small Modular Integral Pressurized Water Reactor

The elimination of soluble boron in the operation of small modular integral pressurized water reactors creates several advantages. Most of these advantages are realized by the core simplification brought on by removing the corrosive effects of soluble boron. Piping, pumps, and tanks associated with soluble boron can be completely eliminated, bringing a significant economic and safety benefit. Additionally, a whole class of accidents related to boron dilution would be eliminated by design, and any loss-of-coolant event would not be affected by the presence of soluble boron. However, removing soluble boron creates its own set of specific challenges that must be overcome. In traditional pressurized water reactors, soluble boron is used in conjunction with burnable poisons to suppress excess initial reactivity. Since boron is diluted in the coolant, its presence is felt uniformly throughout the core, and thus it uniformly reduces the excess initial reactivity. In any boron-free design, an acceptable alternative to boron must be found through the use of the other two mechanisms for reactivity control: burnable poisons and control rods. However, both methods pose challenges. Control rods are actively controlled but are discrete absorbers, locally impacting the core where they are inserted. Since they are inserted from the top of the core, their presence negatively impacts the axial neutron flux profile. This axial flux imbalance creates undesirable peaking factors, leading to reduced operating margins. Thus, the main challenge in any boron-free design concerns excess reactivity suppression and active reactivity control while maintaining a proper axial flux profile and reduced peaking factors. This paper demonstrates that an advanced control rod algorithm with multiple control rod banks can be used for this purpose to satisfy the criteria for success.

A Stylized Three-Dimensional Pressurized Water Reactor Benchmark Problem with UO2 and MOX Fuel

This paper fully describes a whole-core benchmark problem based on a small two-loop pressurized water reactor composed of UO2 and MOX fuel assemblies. The specification includes heterogeneity at both the assembly and the core levels. The geometry and material compositions are fully described, and multigroup material cross-section libraries are provided in the two-group, four-group, and eight-group formats. Detailed Monte Carlo reference solutions including core eigenvalue, assembly-averaged fission distribution, and selected fuel pin fission density distributions are presented for benchmarking diffusion and transport methods. Three different core configurations are presented in this paper, namely, all-rods-out, all-rods-in, and some-rods-in.

Netronic Design of Small Long-Life PWR Using Thorium Cycle

A small long-life core loaded with thorium fuel and 231Pa as burnable poison material has been performed in Pressurized Water Reactor (PWR). Thorium cycle fuel has higher conversion ratio in the thermal spectrum domain and lower reactivity swing than the Uranium-Plutonium cycle fuel. 231Pa have very large capture cross section that can pressed reactivity in the beginning of life. The neutronic analysis result of infinite cell calculation shows that mixed nitride is better than oxide and carbide in thorium fuel system. In the present study we consider thorium nitride system with 3 ~ 8 % 233U percentage and 0.2~ 7% 231Pa as fuel for small PWR and can be burn up for the long time. The purpose of the study is to optimize the design of 350MWt PWR which can be operated without refueling in 10 years The core was designed by cylindrical two-dimension R-Z (radial and axial). The multigroup diffusion and Burn-up analysis was performed by SRAC-CITATION code using libraries based on JENDL 3.2. By using this concept, small PWR can be designed for long time operation with reduced excess reactivity until under 1 % and flatted power distribution during its operation.

Assessment of TAPINS and TASS/SMR codes and application to overpower transients in REX-10

This paper presents numerical studies to assess the applicability of the TAPINS and the TASS/SMR codes for an integral reactor on natural circulation and to evaluate the new reactor design and the transient responses of REX-10 during overpower transients. REX-10 is a small pressurized water reactor with integral design which aims at providing the small-scale electricity generation and the nuclear district heating. For thermal-hydraulic analyses of an integral PWR such as REX-10, an in-house system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed by Seoul National University. Based on the simple field equations formulated using the momentum integral model, TAPINS consists of component models for the core, the helical-coil steam generator, and the steam-gas pressurizer. In this study, the design assessment and transient simulations of REX-10 are carried out using TAPINS, and TASS/SMR selected as another thermalhydraulic code for independent calculations. As preliminary analyses to assess the applicability of these two codes to REX-10, a series of tests at RTF (REX-10 Test Facility) are simulated and the calculation results are compared with the data. Thereafter the prototypical REX-10 is modeled by TAPINS and TASS/SMR. The major design parameters of REX-10 are identified through the steady-state calculation, and the transient responses to overpower transients caused by a reactivity insertion accident and an increase in feedwater event are investigated. The thermal-hydraulic behavior and safety performance of REX-10 are evaluated from the prediction results.

A STUDY ON MODAL CHARACTERISTICS OF FLOW SKIRT USING EFFECTIVE YOUNG'S MODULUS

Many innovative design features are employed in the reactor vessel internals of SMART, a small integral-type pressurized water reactor, one of which is the flow skirt, which uniformly distributes flow and horizontally restrains the lower part of the core support barrel. This new design requires a comprehensive investigation of vibration characteristics. Therefore, in this study, modal characteristics of flow skirts are investigated with finite element analysis. Specifically, we investigate how the presence of holes, the presence of three rings attached to the flow skirt, and the thickness of the lowest shell effect vibration characteristics. In addition, the fluid effect is addressed, since the flow skirt is submerged in the fluid.

Analyses of the OSU-MASLWR Experimental Test Facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.

Small PWR core design with coated particle based fuel with a novel composition

Abstract A conceptual design study for a small-pressurized water reactor (PWR) core has been presented in this research paper that utilizes carbon coated fuel particles with a novel composition. The utilization of novel tristructural-isotropic (TRISO) fuel particles can completely eliminate the use of soluble boron system and requirement of burnable poison if adequate number of control rods is used. Apart from the simplification of the design by excluding such as complex systems from the design, the utilization of TRISO fuel particles in PWR technology can enhance the design integrity by confining the radioactive fission products within fuel itself during reactor operation. Reactor simulation computer codes WIMS-D/4 and CITATION were used for neutronics design and RELAP5 was used for thermal hydraulic analysis. The designed core is small which has only 25 MWth power out put with more than 2 EFPY operating life. The values of Doppler, moderator and void reactivity coefficients were found negative. The Soluble Boron Free (SBF) and Burnable Poison Free (BPF) concepts are more viable in small and medium size reactors (SMRs) because it makes the plant simpler and uniform burnup can also be achieved throughout the core life.

06/02129 Recursively coupled thermal and mechanical FEM–analysis of lower plenum creep failure experiments: Willschütz, H. -G. et al. Annals of Nuclear Energy, 2006, 33, (2), 126–148.

A conceptual design study for a small-pressurized water reactor (PWR) core has been presented in this research paper that utilizes carbon coated fuel particles with a novel composition. The utilization of novel tristructural-isotropic (TRISO) fuel particles can completely eliminate the use of soluble boron system and requirement of burnable poison if adequate number of control rods is used. Apart from the simplification of the design by excluding such as complex systems from the design, the utilization of TRISO fuel particles in PWR technology can enhance the design integrity by confining the radioactive fission products within fuel itself during reactor operation. Reactor simulation computer codes WIMS-D/4 and CITATION were used for neutronics design and RELAP5 was used for thermal hydraulic analysis. The designed core is small which has only 25 MWth power out put with more than 2 EFPY operating life. The values of Doppler, moderator and void reactivity coefficients were found negative. The Soluble Boron Free (SBF) and Burnable Poison Free (BPF) concepts are more viable in small and medium size reactors (SMRs) because it makes the plant simpler and uniform burnup can also be achieved throughout the core life.

Small PWRs using coated particle fuel for district heating, PFPWR50

It has been said that nuclear energy is an important option for especially developing countries to satisfy their increasing energy demand. However, it will be difficult to deploy first of a kind nuclear power plant in developing countries because extensive safety demonstration has to be conducted in industrialized countries. On the other hand, it will be essential to present rigid proof of reliable operational experience to develop proper understanding of the safety features of new reactor systems among the people around the demonstration plant sites. One of the ways to solve the issue is to integrate existing technologies supported by a great deal of data and experience into a new reactor design. Based on the consideration, a small-sized district heating reactor system based on the pressurized water reactor (PWR) technologies combined with the fuel concept of high temperature gas cooled reactors (HTGRs) has been studied. The purpose of the combination of these two existing concepts is to take the best advantages of both excellent operational experience of PWRs and the integrity of HTGR fuel, coated particle fuel, against fission products release even at high temperature. We expect that this approach will help create a breakthrough to the current stagnation of nuclear power deployment.

ICONE11-36579 PROSPECTS FOR NUCLEAR WASTE TRANSMUTATION USING HIGH FLUX SUBCRITICAL REACTORS

A design concept and characteristics for a high flux subcritical reactor (HFSR) are presented. By replacing the control rods with photoneutron generators, we could improve safety and perform radioactive waste burning in subcritical reactors that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). The initial design is based on a small PWR, dry fuel cycle, intense resonance neutron source and power monitoring system with in-core gamma-ray detectors, now under development in the US, Korea, Russia and Ukraine. An important aspect of the HFSR is the reactor's modular design. To increase neutron source intensity the module is divided into two zones : a booster with fission electric cells (FEC) and a blanket with spent/depleted fuel. Neutrons are mostly generated in the booster surrounded the targets. A neutron gate (absorber and moderator) imposed between two subcritical zones permits fast neutrons from the booster flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed. However, these in-core applications require somewhat smaller high-brightness neutron generators that are presently available. An inexpensive method of obtaining large neutron fluxes is target-distributed electron accelerators (TDA), in which a FEC electrical field compensates for lost beam energy in the thin photon production targets. The FEC is essentially a high-voltage power source that directly converts the kinetic energy of the fission fragments into electrical potential of about 2 MV. The charge deposited by the electron beam in the target could be used to suppress the flow of secondary electrons across the gap between the FEC electrodes. Although the low and medium energy accelerator accelerators are mainly seen as support equipment for particle physicists, recent advances in space technology have made it possible to use these accelerators as neutron sources for energy generation systems. The advanced design features, based essentially on the studies in dispersive wave and space physics (wave model of observed relativistic phenomena, ultrasonic concept the plant seismic resistance increase, microwave x-ray source/separator), were recently presented at the 5th Sakharov Conference and 7th Wigner Symposium. Significant savings in cost can be realized by exploring this study in areas as diverse as radio astronomy navigation, space propulsion, monochromatic computed tomography and X-ray lithography.

A New Self-Shielding Method Based on a Detailed Cross-Section Representation in the Resolved Energy Domain

The calculation of a dilution cross section {bar {sigma}}{sub e} is the most important step in the self-shielding formalism based on the equivalence principle. If a dilution cross section that accurately characterizes the physical situation can be calculated, it can then be used for calculating the effective resonance integrals and obtaining accurate self-shielded cross sections. A new technique for the calculation of equivalent cross sections based on the formalism of Riemann integration in the resolved energy domain is proposed. This new method is compared to the generalized Stamm`ler method, which is also based on an equivalence principle, for a two-region cylindrical cell and for a small pressurized water reactor assembly in two dimensions. The accuracy of each computing approach is obtained using reference results obtained from a fine-group slowing-down code named CESCOL. It is shown that the proposed method leads to slightly better performance than the generalized Stamm`ler approach.

Generalization of the Stamm’ler Method for the Self-Shielding of Resonant Isotopes in Arbitrary Geometries

The self-shielding treatment of resonant isotopes is currently performed in most lattice codes using the Stamm’ler method on simplified one-dimensional geometries. A generalization of this procedure is proposed for self-shielding calculations over the arbitrary two- and three-dimensional geometries typical of most advanced reactor designs. Numerical results are presented for a simple two-region cylindrical cell and for a small pressurized water reactor assembly exhibiting true two-dimensional behavior. The absorption rates obtained after self-shielding are compared with exact values obtained using an elastic slowing-down calculation where each resonance is modeled individually in the resolved energy domain. It is shown that the generalized Stamm’ler method can be applied without loss of accuracy to multidimensional domains.

Passive Containment Cooling for an Advanced Small Pressurized Water Reactor

Containment cooling using natural convection and radiation is a key passive safety feature studied for a 600-MW(electric) advanced small pressurized water reactor. A free-standing cylindrical steel...