Conventional Pressurized Water Reactor Research Articles

Research on Reactivity-Equivalent Physical Transformation Method for Double Heterogeneity in Pressurized Water Reactors Based on Machine Learning

Traditional computational methods for pressurized water reactors are unable to handle dispersed fuel particles as the double heterogeneity and the direct volumetric homogenization can result in significant errors. In contrast, reactivity-equivalent physical transformation techniques offer high precision for addressing the double heterogeneity introduced by dispersed fuel particles. This approach converts the double heterogeneity problem into a single heterogeneity problem, which is then subsequently investigated by using the conventional pressurized water reactor computational procedure. However, it is currently empirical and takes a lot of time to obtain the right k∞. In this paper, we train the RPT model by using the existing dataset of plate-dispersed fuel and rod-dispersed fuel by a machine learning method based on a linear regression model, and we then use the new data to make predictions and derive the corresponding similarity ratios. The burnup verification, density verification, fission rate verification, and neutron energy spectrum analysis are calculated through the OpenMC program. For plate-type fuel elements, the method maintains an accuracy within 200 pcm during depletion, with deviations in the 235U density and 235U fission rate within 0.1% and neutron energy spectrum errors within 6%. For rod-type fuel elements, the method maintains an accuracy within 100 pcm during depletion, with deviations in 235U and 239Pu density within 1.5% and neutron energy spectrum errors within 1%. The numerical validation indicates that the reactivity-equivalent physical transformation method based on the linear regression model not only greatly improves the computational efficiency, but also ensures a very high accuracy to deal with double heterogeneity in nuclear reactors.

Evaluation of proliferation resistance of SMRs with COMPRE methodology assisted by Pu production

The proliferation resistance (PR) of several small modular reactors (SMR) is investigated in terms of material characteristic and other extrinsic parameters with COMPRE methodology developed by KINAC. In this work, plutonium (Pu) production amount and isotopic yield are newly introduced as key metrics in material characteristic. Their functionalization is performed with a linear regression method based on sparse core depletion data. Nevertheless, a clear discrepancy is observed in Pu buildup and its quality depending on SMR type and core burnup status. Some of the non-light water SMRs exhibit a higher proliferation concern and their PR score trends are clearly different compared to conventional pressurized water reactors (PWR). Other extrinsic parameters are evaluated based on development status of representative SMRs. At present, liquid metal cooled-fast reactor (LMFR) and molten salt reactor (MSR) receive low scores due to their limited technical maturity in reactor design and related I&C equipment. We conclude that distinct safeguards approaches, suitable for a wide range of SMR designs should be developed for commercial use in the near future.

Flexible operation of nuclear hybrid energy systems for load following and water desalination

Nuclear hybrid energy systems (NHES) have the potential to provide dependable and emission-free electricity to the grid while also increasing the flexibility and reliability of the electrical grid. Molten salt reactor (MSR) technology can provide consistent, carbon-free electricity while also increasing efficiency, security, and sustainability and reducing nuclear waste. This study investigates the integration of Molten Salt Reactors (MSR) and conventional Pressurized Water Reactors (PWR) with desalination technologies: Direct Contact Membrane Distillation (DCMD), Multi-Stage Flash Distillation (MSFD), and Reverse Osmosis (RO). Dynamic first-principles models were developed and tested using real grid data from the New York Independent System Operator. The results demonstrate that nuclear power is capable of flexibly responding to changing grid demand while simultaneously producing clean water, particularly during periods of low electricity demand. The MSR-RO system was found to be the most efficient in electricity generation and water production, and all hybrid systems reduced CO2 emissions by 356,000 to 682,000 tons annually. Economic analysis reveals that nuclear desalination technologies are cost-competitive with conventional systems, especially when paired with RO. These findings confirm the technical feasibility and environmental benefits of nuclear hybrid systems for sustainable electricity and water production.

Performance evaluation of nuclear fuel in a reactor based on the CAREM 25

Nuclear energy has been proven as a viable option for strengthening a power grid, and the expansion of the Brazilian nuclear programme is becoming a reality in light of new reactor projects currently in the licensing and/or construction stage in the country. The object of study in this paper is a small modular reactor (SMR) based on the CAREM 25 Reactor designed in Argentina that represents an innovation in nuclear reactor research in Latin America and belongs to the category of advanced reactors. It is a reduced-size Pressurized Water Reactor (PWR) with integrated primary and secondary circuits, passive and redundant safety systems, light water cooling, and a power output of 25 MW. In this study, the behaviour of certain safety parameters related to a nuclear fuel rod operating at high burnup is evaluated, considering different types of coatings commonly used in the country. The aspects addressed are related to plant operational safety and fuel integrity, such as fuel centreline temperature, cladding surface temperature, oxide layer thickness on the rod surface, and hydrogen concentration in the cladding. The results obtained show that the tool used for simulating the performance of fuel rods in conventional PWR reactors, such as the FRAPCON line, can be applied to smaller modular reactors as well, along with the feasibility of using well-established coatings in the country.

What Would Happen to a Fusion Power Plant in the Case of Plasma Disruption? A Dynamic Model Simulation

It is a widespread view in the fusion community that steady-state, water-cooled fusion power plants can utilize the power generation systems of conventional pressurized water reactor (PWR) fission plants as is. However, what would happen to a fusion power plant in the case of plasma disruption? The authors constructed a dynamic simulation model of a water-cooled ceramic breeder blanket fusion power plant model on Modelica language [300.0-MW(electric) electrical output/1138-MW(thermal) fusion output] and evaluated the applicability of a PWR power generation system. Simulation results suggest that while the PWR system would function as intended during steady-state operation, the conventional system may not be able to cope with a sudden loss of energy influx in the event of plasma disruption without modification: The PWR system’s steam generator experienced a water overflow in less than 150 s from the plasma disruption.

Investigation on long-term daily load follow operation capability of soluble boron-free SMR

Soluble boron-free reactors inherently have larger reactivity feedback than conventional pressurized water reactors during power transients. The purpose of this study is to verify the capability of the long-term load follow operation with only control rods for the soluble boron-free small modular reactors. The effects of the long-term load follow operation on burnup distribution, power distribution, reactivity balance and spatial xenon stability are investigated throughout the whole cycle.The reactivity balance results show that the control rod design of the soluble boron-free SMR successfully manages the reactivity changes introduced by power and xenon changes during the daily load follow operation. The spatial xenon stability analysis demonstrates that the SMR model has enough capability to suppress the axial and quadrant xenon oscillation without a second peak during the load follow operation. The load follow simulation results showed that safety parameters including peaking factor, axial offset and shutdown margin satisfy design limits.

Special Section: Selected Papers From the Ninth International Symposium on Supercritical Water-Cooled Reactors

Special Section: Selected Papers From the Ninth International Symposium on Supercritical Water-Cooled Reactors

Whole-core analyses on recriticality of conventional high power pressurized water reactor in Korea during early phase of severe accidents

In this paper, first, MELCOR calculations for a conventional high power pressurized water reactor are performed to obtain the geometrical configurations and remaining fraction of the fission products in the degraded reactor core during the early phase of a severe accident. Then, based on the coupling between MELCOR and Serpent 2 code, sensitivity analyses on the criticality are performed for various water levels, burnup, and release of fission products with whole-core modelling on the degraded reactor. Sub-critical boron concentrations are obtained on the configuration of the degraded reactor core showing the largest criticality. The sub-critical boron concentrations are compared to critical boron concentrations of the intact reactor core to quantify additional boron concentrations required to make the reactor sub-critical during the early phase of the severe accident. Effects of control rod degradation and fission product release are also presented in terms of boron concentrations from the results of sensitivity analyses.

Three-Dimensional Pin-Resolved Transient Diffusion Analysis of PWR Core by the Hybrid Coarse-Mesh Finite Difference Algorithm

This paper presents the methodology and performance of the Hybrid Coarse-Mesh Finite Difference (HCMFD) algorithm for transient pinwise analyses of three-dimensional (3-D) pressurized water reactor (PWR) problems. The time-dependent neutron diffusion equations and their applications in two steps of the HCMFD algorithm, i.e., local and global iterations, are introduced in detail. Taking into account the characteristics of the local-global nonlinear HCMFD iterations, an optimization strategy to minimize the computing time of the transient HCMFD calculation is established by focusing on the balance between the number of local and global calculations. Based on the optimization strategy, the actual computational performance of the transient HCMFD algorithm, in view of both computing time and accuracy, is evaluated for the core of a big-sized conventional PWR in this work. To demonstrate the effectiveness of the optimized iteration strategy, various slow and fast transients including a rod ejection transient are simulated by the transient HCMFD algorithm. It is clearly shown that a 3-D pin-resolved whole-core transient solution for a big PWR can be obtained in a reasonably short computing time by the transient 3-D HCMFD algorithm.

Reactor Monte Carlo (RMC) model validation and verification in compare with MCNP for plate-type reactor

Traditional neutronics calculations and simulation behavior of plate type research reactors pose a challenge comparing with the conventional pressurized water reactors. This is due to its complex geometry, characteristics, heterogeneity, and neutron leakage that lead to modeling difficulties. The Reactor Monte Carlo code (RMC) is employed in stochastic approach for the neutronic analysis in the plate type research reactor. The goal of this report is to examine the capability and validity of the model built using RMC and compare the results with a well-known Monte Carlo code such as MCNP. The results between RMC and MCNP are in a good agreement for effective multiplications factor (keff), radial power peaking factor, total power peaking factor, and integral and differential control rods worth.

Physics conditions of passive autonomous frequency control operation in conventional large-size PWRs

The non-base load operation of nuclear power plants is expected to be unavoidable in future due to the increasing shares of the intermittent renewables. In this paper, the physics conditions required for a passively autonomous frequency control operation (PAFO) in a 3400 MWt conventional pressurized water reactor (PWR) design is investigated. The PAFO scheme allows a PWR to passively achieve the requested power maneuverability for primary and secondary frequency regulations on the electrical grid. This study is carried out for a conventional coolant system with a typical critical boron concentration, which yields a less negative coolant temperature coefficient (CTC) than that in a soluble-boron-free (SBF) core. The PAFO scheme is expected to be more challenging in the conventional PWRs than in an SBF small modular reactor, particularly for the secondary frequency regulation. This is mainly due to the weaker coolant reactivity feedback and the higher fuel temperature, which yields a stronger Doppler reactivity feedback during the passive power transients. Numerical simulations of PAFO are performed, for typical power ranges and power ramping rates required for frequency regulations, using a lumped PWR model that is solved by an in-house Fortran-95 computer code. In conventional PWRs, the progressive dilution of the soluble-boron throughout the cycle and the fuel depletion vary the reactivity coefficients. Therefore, we perform sensitivity analyses on CTC, FTC, and Xe worth to assess the feasibility of PAFO throughout the fuel cycle. In addition, sensitivity analysis on the fuel temperature is done to study its role in altering the reactor performance during PAFO.

Reactor physics analysis of a new Accident Tolerant Fuel called Fuel-in-Fibers

Fuel-in-Fibers (F-in-F) is an Accident Tolerant Fuel (ATF) concept aimed at significantly increasing the fission product retention capability of nuclear fuel at high temperatures, similar to another ATF concept, Fully Ceramic Microencapsulated Fuel (FCM). FCM, however, when implemented in a conventional Pressurized Water Reactor (PWR) fuel geometry, is not able to maintain an 18-month fuel cycle length and remain below the proliferation enrichment limit of 20 w/o 235U using the standard cycle reload strategy. The F-in-F enrichment is decreased by putting the fuel in a cylindrical fiber instead of a spherical particle in order to increase the volume fraction of fuel in the pellet. The fibers are created with a direct manufacturing process that allows for the minimization of the volume occupied by the SiC matrix and the maximization of uranium loading via direct deposition of high density fuels like uranium nitride (UN). The full core reactor physics simulation showed that when the F-in-F design used UN as a fuel, it only required enrichments of 6% to maintain an 18-month fuel cycle length. Based on a full core analysis when compared to the current reference fuel for a standard 4-loop Westinghouse PWR, F-in-F has a similar steady state thermal hydraulic and neutronic performance.

Developing a Full-Core MCNP6 and RELAP5 Model of the European Pressurized Reactor Using NWURCS

The European pressurized reactor (EPR) is classified as a Generation III+ reactor. It differs from a conventional pressurized water reactor in many aspects, one of which is the core design. This evolutionary reactor lends itself to new fuel designs, such as thorium-based fuels. To perform new design calculations, a base case model needs to be established because the detailed models that are currently available are either proprietary or regulated. This paper therefore presents such a model based on the Monte Carlo method. This method is a valuable component of reactor neutronic calculations because geometry and materials can be accurately modeled.We modeled a full core of the EPR using MCNP6, in which the individual fuel pin geometry and material definitions were used together with radial and axial temperature characterization based on fuel assemblies considered as nodes. Data for both the neutronic and thermal-hydraulic models were mainly obtained from the U.S. EPR Final Safety Analysis Report (FSAR) [Rev. 5, AREVA (2013)].The neutronic and some thermal-hydraulic results were compared with data from the EPR FSAR. The following core neutronic parameters compared well with the FSAR data: the boron worth, axial flux distribution, neutron flux spectrum, reactivity coefficients, and control rod worth. However, the delayed neutron fraction showed a somewhat larger difference compared to the FSAR. Given this verification with the FSAR, confidence in the MCNP6 EPR model was therefore established. The model that we have developed serves as the basis for the follow-on study of introducing thorium in the EPR core.

Self-pressurization analysis of the natural circulation integral nuclear reactor using a new dynamic model

Abstract Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established.

Investigation of thermal hydraulic behavior of SBLOCA tests in SMART-ITL facility

An experimental investigation of the thermal hydraulic behavior of small break loss of coolant accident (SBLOCA) tests in an integral test loop (ITL) of a system-integrated modular advanced reactor (SMART) was performed, especially on the pressurizer safety valve (PSV) line and safety injection (SI) line breaks. Compared to typical phases of SBLOCAs in conventional and advanced active pressurized water reactors (PWRs), four sequential phases were identified: a blowdown to the upper downcomer (UDC) under saturation conditions, a pressure plateau under forced circulation, boil-off after the reactor coolant pump (RCP) trip, and core level restoration after the safety injection tank (SIT) injection or long-term cooling. The pressure plateau was deeply dependent under UDC saturation conditions and there were mass transfers among the reactor vessel (RV) regions during the pressure plateau. In the boil-off phase, a short reverse flow occurred from the lower dowoncomer (LDC) to the primary side of steam generator (SGP) owing to the occurrence of vaporization in the LDC region. The core level restoration is mainly dependent on the SIT injections. In the secondary system, the fluid conditions on the secondary side of steam generator (SGS) outlets were changed during the tests from superheat to saturation, from saturation to subcooling, and from subcooling to superheat. In the passive safety injection system (PSIS), there was a short reverse flow in the pressure balance lines (PBLs) of the core make-up tanks (CMTs) just after the break, and sufficient injection flowrates of the CMTs were achieved after the partial clearing/blocking of the PBL. In the case of a SIT actuation signal, duration times for the hydraulic equilibrium in pressure among the CMTs, SITs, and RV were needed.

Fuel cycle analysis of Advanced Burner Reactor with breed-and-burn thorium blanket

The Seed-and-Blanket (S&B) Sodium-cooled Fast Reactor (SFR) core concept was proposed for generating a significant fraction of the core power from thorium fueled breed-and-burn (B&B) blankets without exceeding the presently verified radiation damage constraint of 200 Displacement per Atom (DPA). To make beneficial use of the excess neutrons from fast reactors, the S&B core is designed to have an elongated TRU transmuting (or “TRU burner”) seed from which over 20% of the fission neutrons leak into a subcritical thorium blanket that radially surrounds the seed. The seed fuel is recycled while the blanket operates in a once-through breed-and-burn (B&B) mode. The objective of this paper is to compare the fuel cycle performance of the S&B reactor against an Advanced Burner Reactor (ABR) and a conventional Pressurized Water Reactor (PWR). For the fast reactors (SFR: ABR and S&B) the fuel cycle performance is evaluated based on a 2-stage PWR-SFR energy system while the reference nuclear system is made of once-through PWRs.It was found that relative to the ABR, the S&B core has a lower fuel cycle cost, higher capacity factor, and comparable short-term radioactivity. The discharged seed fuel from the S&B core features lower fissile Pu-to-Pu ratio, higher 238Pu-to-Pu ratio, higher specific plutonium decay heat, higher spontaneous fission rate, and lower overall material attractiveness for weapon use. Due to the significant amount of 233U discharged from the breed-and-burn thorium fueled blankets, the S&B core has much higher long-term radioactivity and radiotoxicity. Since the thorium fueled blanket operates in the breed-and-burn mode and requires no fuel reprocessing, the discharged blanket fuel is unattractive for weapons application.Compared with a PWR, the S&B core has a lower fuel cycle cost, much lower short-term radioactivity and radiotoxicity but higher long-term values, and higher proliferation resistance for the discharged plutonium. The natural uranium utilization of the 2-stage PWR-S&B system is approximately 60% higher than that of present PWRs; it is few percent higher than that of the 2-stage PWR-ABR system. Approximately 7% of the thorium fed to the blanket is converted into energy, which makes the thorium fuel utilization approximately 12 times the utilization of natural uranium in PWRs.A comprehensive fuel cycle evaluation performed with the methodology developed by the recent U.S Department of Energy’s Nuclear Fuel Cycle Evaluation and Screening campaign concludes that the PWR-S&B system has similar fuel cycle performance characteristics as the PWR-ABR system. The S&B concept may potentially feature improved economics and resource utilization relative to the ABR.

Technology Selection for Offshore Underwater Small Modular Reactors

Abstract This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead–bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Thermo-hydraulic analysis of the supercritical water-cooled reactor core by porous media approach

Super critical water reactors (SCWR) are high efficient version of conventional pressurized water reactor with the benefit of using super critical water as coolant and moderator. Researches and studies on nuclear reactor core are usually subdivided into two major fields, named: thermal–hydraulic and neutronic, in which, precise simulation of reactor behavior in both fields is highly required to ensure the designers that reactor will work in a safe margin. In this study, a thermal–hydraulic analysis of a typical SCWR with three pass core design is performed using a porous media approach. Based on this approach, each fuel assembly was modeled and was divided into a network of lumped regions; each of them was characterized by a volume average parameter. In such manner, while complex geometries are easily defined and dealt with, the thermal–hydraulic parameter and phenomena like friction, shear stress, cross-flows, convective heat transfer and etc. are strictly included in simulations. To validate the applied approach, the numerical analysis and available data in literature were compared and showed a good agreement with those reported by other authors.

Experiments on the Effects of a Spacer Grid in Air-Water Two-Phase Flow

Understanding the effects of spacer grids on the coolant flow through a nuclear reactor core is required for best-estimate design and analysis of the plant. The impact of a spacer grid on two-phase flows is of particular importance because the geometric effects of the grid can alter the two-phase flow structure and, consequently, the mass, momentum, and energy transfer characteristics. Therefore, a scaled separate-effects test facility is constructed to investigate the effects of a spacer grid on the hydrodynamics of air-water two-phase flow through a rod bundle. The test facility is scaled to maintain hydrodynamic and geometric similarity to single- and two-phase flows in a conventional pressurized water reactor and to facilitate detailed local measurements of two-phase flow parameters around the simulant fuel rods with a four-sensor conductivity probe. This paper presents measurements of local time-averaged two-phase flow parameters acquired upstream and downstream of the spacer grid with the conductivity probe in a representative subchannel of a 1×3 rod bundle for eight flow conditions. Characteristic features of the development of the two-phase flow parameters along the test section are discussed.

Feasibility of Fueling PWRs with Fuel Discharged from Breed and Burn Reactors

Abstract The fuel discharged from Breed and Burn (B&B) reactors contains a relatively high content of fissile plutonium (∼10%). This study assesses the feasibility of reducing this fissile content while increasing the fuel utilization by loading the B&B discharged fuel, after reconditioning, into Pressurized Water Reactors (PWR). Two processes are examined for recycling the B&B reactor discharged fuel: the melt–refining process and the AIROX process. Both use processes that can remove a fraction of the fission products and cannot be used for actinide separation. It is found possible to load full PWR core with the reconditioned fuel and operate it to an additional burnup of up to 70,000 MWd/MT IHM (AIROX processed fuel) and 105,000 MWd/MT IHM (melt-refining treated fuel), while maintaining a negative coolant temperature reactivity coefficient. The burnup reactivity swing is significantly smaller than in conventional PWR cores. The B&B reactor fuel is highly proliferation resistant after being discharged from the PWR.