Light Water-cooled Reactor Research Articles

The Reactivity of Hydroxyl Radicals toward Boric Acid as a Function of pH.

Boric acid and its counter-base, borate, are a commonly used buffer pair in many systems where hydroxyl radicals are generated. Boric acid is also used in light water-cooled nuclear reactors to control the excess reactivity of the nuclear fuel. Hydroxyl radicals are generated within the cooling water of the reactor because of intense radiation. The reactivity of the hydroxyl radical toward boric acid has previously been studied, but to the best of our knowledge, only upper limits of the rate constants are available in the literature. In this study, the rate constants for the reaction between the hydroxyl radical and boric acid and its counter-base including several polyborates that form at high boron concentration are determined. The rate constants were determined from competition kinetics using steady-state γ radiolysis and coumarin-3-carboxylic acid as the competing reactant. By varying the pH and accounting for boron speciation, it was possible to determine the rate constant for the different boron species using multilinear regression. The rate constants for boric acid and the counter-base were determined to be 3.6 × 104 and 1.1 × 106 M-1·s-1, respectively, which is very close to the previously determined upper limits of the rate constants. For the polyborate species diborate and tetraborate, the rate constant was determined to be 6.4 × 106 and 6.8 × 106 M-1·s-1, respectively.

Thermal creep in a pre-hydrided Zr1 % Nb nuclear fuel cladding tube

The effect of hydrogen on the thermal creep behaviour of Zr1 %Nb alloy fuel cladding tubes for use in light water-cooled nuclear reactors was investigated. The tubes were hydrogen charged using an alternative method with a hydrogen content in the range 371–656 wppm H. Comparative constant-stress creep tests were carried out at 350 °C, under stress ranging from 150 to 225 MPa, on both non-hydrided and hydrided tubes. These creep tests were followed by a microstructural analysis of the specimens exposed to creep using SEM and TEM. It was found that these creep tests were conducted in the transition zone between the power-law and power-law breakdown creep regions. Hydrogen, either in solid solution (atomic hydrogen) or as precipitated zirconium hydrides, altered the creep behaviour of the hydrided tubes, and the impact of hydrogen depended on the hydrogen content CH. It was observed that under the creep loading conditions considered here with a hydrogen content CH < 450 wppm H, the presence of hydrogen resulted in an increase in the creep rate; however, a higher concentration of hydrogen caused a decrease in the creep rate. The difference in the effects of hydrogen on the creep rate was explained by the different roles of hydrogen in solid solution and in the form of zirconium hydride precipitates.

Scope of conceptual development of Resilient supercritical Light water-cooled reactor (SCWR-R)

The research and development (R&D) of Supercritical Light Water-cooled Reactor (SCWR) has been based on the matured Generation III light water-cooled reactor (LWR) technology and the supercritical fossil fuel-fired power plant (SC-FPP) technology. The plant concept has unique features such as the simple and compact plant system and the high thermal efficiency. Currently, the Japanese concepts are being further developed at Waseda University with focuses on improving resilience of SCWR. This study proposes and presents the overall scope of the new Resilient SCWR (SCWR-R) concept with focuses on two viewpoints: (1) SCWR can utilize variety of different core designs ranging from thermal to fast spectrum for multi purposes (e.g., Pu breeding, Minor Actinide burning) with the same plant system to respond to different fuel cycle needs; (2) severe accident management of SCWR-R can be developed with the combination of IVR and PCV with S/C to minimize the PCV internal contamination by radioactive materials, which should greatly reduce post-severe accident management. To establish the SCWR-R concept, the multi-level physics simulation methodology is currently being developed to evaluate the debris criticality, cooling and radioactive aerosol management during the severe accident.

Numerical analysis of nucleate boiling in rolling rod bundle with tilted axis using two-fluid modeling approach

In this study, an Eulerian-Eulerian two-fluid approach was applied to simulate nucleate boiling in a vertical rolling 2 × 2 rod bundle with a tilted axis. For this purpose, the standard multiphaseEulerFoam solver of OpenFOAM was modified to include the fictitious forces necessary in a non-inertial frame of the reference formulation. High-performance computing was then applied to analyze the effects of oscillation on boiling and the physics of the fluids inside the rod bundle. In particular, the effect of a tilted rolling axis was investigated for the first time. The results indicated that the space-averaged void fraction and surface-averaged heat transfer coefficient decrease in the rolling cases compared with the stationary case. Furthermore, it was found that while the tilt angle has a negligible effect on the space-averaged values, it may have a significant impact on the maximum value of the void fraction. The higher the tilt of the rolling axis, the higher the variation in the maximum field values. The results of this study will further improve our understanding of wall nucleate boiling, which is critical for the design, development, and safety of light water-cooled reactors (LWR).

Design Study of Small Modular Reactor Class Super Fast Reactor Core for In-Vessel Retention

Abstract The small modular reactor (SMR) class core design concept of once-through supercritical light water-cooled reactor (SCWR) with fast neutron spectrum (super fast reactor (FR)) is being developed at Waseda University. For the 300 MWel class design, complete core meltdown may need to be considered in case of a severe accident. This study proposes the new in-vessel retention (IVR) concept of the SMR class super FR (super FR-IVR), which can avoid recriticality even if the whole core relocates to the lower plenum of the reactor pressure vessel (RPV). The core characteristics with a given set of design specifications and criteria are evaluated based on fully coupled neutronics and thermal-hydraulics core burnup calculations. The debris criticality is evaluated based on Monte Carlo based method to consider the RPV lower plenum and debris configurations. The relationships between the 300 MWel class core design with the inner vessel diameter of 2.32 m and the IVR design are revealed. By reducing the operation cycle length from 720 days to 360 days and increasing the core inlet temperature from 280 °C to 370 °C, the required IVR submergence level could be reduced from 2.18 m to 1.38 m, assuming that the debris bed (melt pool) is homogeneous. However, when fully stratified debris configurations are assumed, the required disperser height and the corresponding IVR submergence level may increase to about 2.70 m.

Effect of dpa rate on the temperature regime of void swelling in ion-irradiated pure chromium

Pure chromium is a promising candidate for coating of Zircalloy fuel rods in light water-cooled power reactors to avoid or delay hydrogen generation in accident scenarios. Void swelling of chromium is one possible contributor to coating-interface failure and needs to be studied. The effect of dpa rate on void swelling of Cr was studied using 5 MeV Fe ion irradiation to 15 peak dpa at peak dpa rates of 3.5 × 10−5, 3.5 × 10−4, and 3.5 × 10−3 dpa/s, at six temperatures between 350 °C and 650 °C. The post-transient (steady-state) swelling rate of pure Cr is ∼0.05%/dpa. Swelling in Cr also appeared to start at a much higher swelling rate at very low doses. The observed dependence of peak swelling temperature on dpa rate agrees well with earlier theoretical models and clearly demonstrated the well-known “temperature shift” phenomenon. After determination of the peak swelling temperatures under ion irradiation at three dpa rates we extrapolated downwards to dpa rates characteristic of both sodium-cooled fast reactors and light water-cooled reactors.

Coupled Analysis of Thorium-based Fuels in the High-Performance Light Water Reactor Fuel Assembly

One of the six selected concepts to be part of Generation IV nuclear reactors is the Supercritical Light Water Cooled Reactor. The High-Performance Light Water Reactor (HPLWR) is the European version and it is a very promising design. In recent years, interest in the study of thorium-based fuel cycles has been renewed and its possibilities for current LWRs have been evaluated. The use of thorium-based fuels will be fundamental in the future sustainability of nuclear energy, since in addition to its abundance in nature, thorium has an important group of advantages. In this paper, performance of thorium-based fuels in the typical fuel assembly of the HPLWR reactor is evaluated, using a computational model based on CFD and Monte Carlo codes for the neutronic/thermal-hydraulic coupled analysis. The volumetric power density profiles, coolant temperature profiles, fuel temperature profiles and others are compared with those obtained for standard UO 2 fuel. When the thorium-based fuels are used, the obtained infinite multiplication coefficients are smaller than the value obtained when UO 2 is used, since the 232 Th isotope has a lower contribution to the multiplicative properties of the medium than 238 U. As a result, a difference of approximately 12 000 pcm was observed. The results verified that the HPLWR is a thermal reactor with a hard spectrum. There are no notable changes in the neutron spectrum if the mass fraction of thorium is slightly varied. With coupled analysis, the potential benefits of the utilization of thorium-based fuels were verified. Moreover, a significant temperature decrease by 136 K on the center line of the fuel elements was observed. When the mass fraction of thorium increases in the oxides mixture, the weighted average temperature on the fuel elements decreases .

Transition cycle analysis of light water-cooled SMR core loaded with MOX (TRU) and FCM (TRU) fueled PWR assemblies

In this study, a light water-cooled small modular reactor (SMR) core loaded with special fuel assemblies composed of mixed oxide (MOX) fuel rods using UO2-TRUO2 and fully ceramic micro-encapsulated (FCM) fuel rods using TRUO2 was suggested and analyzed for transuranic (TRU) transmutation as an alternative option before the commercialization of sodium-cooled fast reactor (SFR). Especially, the FCM rods were loaded with only TRUO2 for deep burning of TRU. This study performed neutronic analysis of the transition cycle cores with the new fuel assemblies. In addition to the neutronic analysis, detailed analyses on two important issues for the light water-cooled reactor (LWR) core loaded with TRU were performed in fuel assembly level calculation. First, an effective burnable absorber design was searched to mitigate the degradation in effectiveness of the burnable absorber caused by the hardened neutron spectrum. Secondly, a reactivity decomposition analysis was performed for understanding quantitative nuclide-wise and reaction-wise contributions to the void reactivity which is a key safety parameter of TRU-bearing fuel assemblies. The analysis on the TRU mass flow through the transition cycles showed that a high net TRU consumption rate of 14.7% can be achieved at the equilibrium cycle.

A MESOSCOPIC OXIDE FUEL CLUSTERING AND ITS GLOBAL PERFORMANCE

Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppler broadening, moderation ratio, and, sometimes, neutron gas temperature etc.) and on variations of thermal-physics parameters (temperature distributions, fuel elongation and moderator density). Usually, in a rough reactor analysis one ignores the very details of temperature distributions largely staying in a frame of so-called adiabatic assumptions (when temperature and density distribution are changing in sync keeping given spatial shapes). In majority of practical applications the radially distributed temperature fields are represented as monotonically smeared ones as if fissile and other materials are homogeneously mixed. Moreover, no one measurement technique allows counting precise correlation between reactivity feedback and in-pellet temperature and materials space-time distributions. However, if fuel is made of Mixed Oxide Plutonium-Uranium compound the behavior of Light Water Reactor would be impacted by an appearance of Pu-rich agglomerates that could be large enough to change physical processes. In such case the fuel reacts on power and temperature variations no more as a homogeneous but a heterogeneous media (on a mesoscopic scale, of course). It leads to changes in a fission product distributions, a fission gas release and, even, to an appearance of multiple components in a Fuel Temperature Coefficient and in a Power Reactivity feedback. These components would depend non-linearly on power, power rate and on some details of a heat transfer. This paper is the only first step of a broad research program where we are estimating the relevant phenomena just by an order of magnitude.

Startup Thermal Analysis of a Supercritical-Pressure Light Water-Cooled Reactor CSR1000

Supercritical-pressure light water-cooled reactors (SCWR) are the only water cooled reactor types in Generation IV nuclear energy systems. Startup systems, and their associated startup characteristic analyses, are important components of the SCWR design. To analyze the entire startup system, we propose a wall heat transfer model in a paper, based on the results from a supercritical transient analysis code named SCTRAN developed by Xi’an Jiao tong Tong University. In this work, we propose a new heat transfer mode selection process. Additionally, the most appropriate heat transfer coefficient selection method is chosen from existing state-of-the-art methods. Within the model development section of the work, we solve the problem of discontinuous heat transfer coefficients in the logic transformation step. When the pressure is greater than 19 Mpa, a look-up table method is used to obtain the heat transfer coefficients with the best prediction accuracy across the critical region. Then, we describe a control strategy for the startup process that includes a description of the control objects for coolant flow rate, heat-exchange outlet temperature, system pressure, core thermal power, steam drum water-level and the once-through direct cycle loop inlet temperature. Different control schemes are set-up according to different control objectives of the startup phases. Based on CSR1000 reactor, an analytical model, which includes a circulation loop and once-through direct cycle loop is established, and four startup processes, with control systems, are proposed. The calculation results show that the thermal parameters of the circulation loop and the once-through direct cycle meets all expectations. The maximum cladding surface temperature remains below the limit temperature of 650℃. The feasibility of the startup scheme and the security of the startup process are verified.

Grain growth in uranium nitride prepared by spark plasma sintering

Uranium mononitride (UN) has long been considered a potential high density, high performance fuel candidate for light water reactor (LWR) and fast reactor (FR) applications. However, deployability of this fuel has been limited by the notable resistance to sintering and subsequent difficulty in producing a desirable microstructure, the high costs associated with 15N enrichment, as well as the known proclivity to oxidation and interaction with steam. In this study, the stimulation of grain growth in UN pellets sintered using SPS has been investigated. The results reveal that by using SPS and controlling temperature, time, and holding pressure, grain growth can be stimulated and controlled to produce a material featuring both a desired porosity and grain size, at least within the range of interest for nuclear fuel candidates. Grain sizes up to 31 μm were obtained using temperatures of 1650 °C and hold times of 15 min. Evaluation by EBSD reveal grain rotation and coalescence as the dominant mechanism in grain growth, which is suppressed by the application of higher external pressure. Moreover, complete closure of the porosity of the material was observed at relative densities of 96% TD, resulting in a material with sufficient porosity to accommodate LWR burnup. These results indicate that a method exists for the economic fabrication of an 15N-bearing uranium mononitride fuel with favorable microstructural characteristics compatible with use in a light water-cooled nuclear reactor.

Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1eV and 105eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

Single pass core design for a Super Fast Reactor

Single pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward coolant flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern significantly simplifies the core concept in terms of upper core structure, coolant flow scheme as well as refueling procedure. In single pass core design, supercritical-pressure water at approximately 25.0MPa enters the core at 280°C and heated up in one through upward flow to the average outlet temperature at 500°C. Great coolant density change in vertical direction will cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO2 also introduces great power increase in blanket assemblies during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of blanket assemblies to mitigate the power change in blanket assembly and to increase the coolant outlet. With exclusive refueling and shuffling scheme as well as partial length control rod design, neutronics & TH coupled calculation shows satisfactory results that can fulfil the requirement of design for both 500°C outlet temperature and negative coolant void reactivity.

Analysis of anticipated transient without scram of a Super Fast Reactor with single flow pass core

Anticipated transient without scram (ATWS) of the supercritical-pressure light water-cooled fast reactor (Super FR) with single flow pass core is analyzed to clarify its safety characteristics. Abnormal reactivity due to loss of feedwater heating and uncontrolled withdrawal of a control rod is mitigated by density and Doppler reactivity feedback. Increase of the coolant density at the upstream of the fuel channel leads the core flow rate to decrease which in turn decreases the power especially due to the density reactivity feedback. Meanwhile, ATWS accidents which are followed by trip of both RCPs are important. Decreased feedwater flow rate leads directly to a decrease of core flow rate in all fuel assemblies. If the ADS is not used, the MCST increases rapidly due to high power density feature of the Super FR. Coolant expansion effect does not appear in the single flow pass core of Super FR while it appears in the second pass of the two-flow pass core leading to mild increase of MCST of the second pass. However, signal of flow low level 3 (6%) is provided for the ADS actuation and the MCST of each fuel channel is effective to be mitigated due to large core coolant flow rate induced by the ADS operation. The power is also decreased by density and Doppler reactivity feedbacks. MCST and core pressure of all the analyzed events satisfy the criteria.

Various startup system designs of HPLWR and their thermal analysis

This paper summarizes the results of various startup system designs and their thermal analysis of the high performance light water reactor (HPLWR) which is the European version of the various supercritical water cooled reactor proposals. In order to study the thermal-hydraulic characteristics of the HPLWR core, a simplified axial one-dimensional (1D) single channel model is developed, which consists of fuel, cladding, coolant and moderator. The model is verified by the related results of Seppälä (2008). Both constant pressure startup systems and sliding pressure startup systems of HPLWR are presented. In constant pressure startup system, the reactor starts at supercritical pressure. It appears that compared with other SCWR designs, the weight of the component required for constant pressure startup of HPLWR is medium and reasonable. Constant pressure startup systems are found feasible from thermal analysis. And for sliding pressure startup, the reactor starts at subcritical pressure. The adequate core power of 25% with 28% flow rate and a feedwater temperature of 280°C are determined during pressurization phase. The thermal analysis results show that the sliding pressure startup systems for HPLWR are also feasible. Considering the same flow rate as the supercritical-pressure light water-cooled fast reactor (SCFR), the component weight required is reduced in HPLWR.

Core design for super fast reactor with all upward flow core cooling

A 1000MWe level supercritical-pressure light water-cooled fast reactor (Super FR) is proposed in this paper. The newly improved features include two flow paths which both applies upward flow pattern. Compact lattice of MOX fuel pins are arranged in hexagonal seed assembly while blanket assembly comprised of depleted UO2 fuel pins wrapped with solid moderator ZrH layer for negative void reactivity. This core design includes 162 seed assemblies and 73 blanket assemblies with 1.7cm of solid moderation layer of ZrH, the core active height is 3.6m. The axial coolant density distribution of the core is far from uniform due to the all upward flow pattern, however, by applying axial Pu enrichment zoning and new loading and refueling scheme, flattened power distribution is obtained and all the design criteria as well as limitations have been fulfilled with significant delivery of high power density up to 237kw/L, 500°C average core outlet temperature, both whole core and local negative void reactivity.

Total loss of flow accident characteristics of Super FR with new coolant flow

Total loss of flow accident characteristics of the supercritical-pressure light water-cooled fast reactor (Super FR) with all upward flow core cooling in two paths is investigated at supercritical pressure. Upward flow cooling has advantage that flow stagnation does not occur at total loss of coolant flow accident due to the buoyancy of the coolant. It is found that maximum cladding surface temperature (MCST) is sensitive to the volume of the gap between two paths the. By open the automatic depressurization system (ADS), depressurization will induce flow in the RPV, and it can improve the MCST in this accident.

A Perspective on Small Reactor Licensing and Implementation

A great deal of interest has developed recently in the implementation of small reactors in the United States and abroad. Small reactors may offer a significant number of advantages over larger reactors. The diversity of size, design, configuration, and construction features and their planned utilization for nonelectrical power applications as well as traditional power applications pose significant challenges to the current U.S. Nuclear Regulatory Commission (NRC) regulatory structure. The current structure is geared toward nontransportable, commercial, electrical power-producing, light water-cooled reactors utilizing traditional nuclear fuel designs. The NRC is currently engaged in a number of preapplication discussions concerning small reactor designs encompassing three distinctively different technologies. These are integral light water reactors, high-temperature gas-cooled reactors, and liquid metal-cooled reactors. Light water reactor technology-based power generation small reactors will fit best in the current NRC regulatory framework.In response to the anticipated licensing workload, the NRC has implemented organizational changes and has increased its focus in areas supporting the licensing of small reactors. Although the licensing of small reactors has to comply with the requirements imposed by the Atomic Energy Act and the National Environmental Policy Act, there are significant design differences among the various proposed small reactors and the currently licensed reactor designs that result in a number of issues that need to be resolved to properly comply with these statutory requirements. Given the diversity of small reactor designs, a regulatory structure that provides licensing flexibility combined with the required degree of safety assurance would be needed. This is likely to involve a risk-informed and technology-neutral regulatory approach.

Development of a plate-type fuel model for the neutronics and thermal-hydraulics coupled code – SIMMER-III – and its application to the analyses of SPERT

SIMMER-III, a neutronics and thermal-hydraulics coupled code, was originally developed for core disruptive accident analyses of liquid metal cooled fast reactors. Due to its versatility in investigating scenarios of core disruption, the code has also been extended to the simulation of transients in thermal neutron systems such as the criticality accident at the JCO fuel fabrication plant, and, in recent years, applied to water-moderated thermal research reactor transient studies, too. Originally, SIMMER considered only cylindrical fuel pin geometry. Therefore, implementation of a plate-type fuel model to the SIMMER-III code is of importance for the analysis of research reactors adopting this kind of fuel. Furthermore, validation of the SIMMER-III modeling of light water-cooled thermal reactor reactivity initiated transients is of necessity. This paper presents the work carried out on the SIMMER-III code in the framework of a KIT and IRSN joint activity aimed at providing the code with experimental reactor transient study capabilities. The first step of the job was the implementation of a new fuel model in SIMMER-III. Verification on this new model indicates that it can well simulate the steady-state temperature profile in the fuel. Secondly, three cases with the shortest reactor periods of 5.0 ms, 4.6 ms and 3.2 ms among the Special Power Excursion Reactor Tests (SPERT) performed in the SPERT I D-12/25 facility have been simulated. Comparison of the results between the SIMMER-III simulation and the reported SPERT results indicates that although there is space for further improvement on the modeling of negative feedback mechanisms, with this plate-type fuel model SIMMER-III can well represent the transient phenomena of reactivity driven power excursion.

Numerical Analysis of Heat Transfer Test of Supercritical Water in a Tube Using the Three-Dimensional Two-Fluid Model Code

The three-dimensional two-fluid model analysis code ACE-3D is developed in Japan Atomic Energy Agency for the thermal design procedure on two-phase flow thermal-hydraulics of light water-cooled reactors. In order to perform thermal hydraulic analysis of SCWR, ACE-3D is enhanced to supercritical pressure region. As a result, it is confirmed that transient change in subcritical and supercritical pressure region can be simulated smoothly using ACE-3D, that ACE-3D can predict the results of the past heat transfer experiment in the supercritical pressure condition, and that introduction of thermal conductivity effect of the wall restrains fluctuation of wall temperature.