Conventional Light Water Reactors Research Articles

Material attractiveness of irradiated fuel salts from the Seaborg Compact Molten Salt Reactor

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

Initial Verification and Validation of a New CASMO5 JENDL-5 Nuclear Data Library for Typical LWR Applications

A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, the range of data, such as the number of nuclides, has been dramatically expanded for general-purpose applications. At the same time, many essential nuclides for conventional light water reactor (LWR) analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5–based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1–based library as possible with JENDL-5. This study performs initial verification and validation of the new library under typical LWR conditions. Verifications were performed based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark phase III-C, and the calculated kinf and fuel compositions of the boiling water reactor fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, a tank-type critical assembly critical experiment and Takahama-3 postirradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1–based library in predictions of reactivity and fuel compositions for typical LWR systems.

High volume packing fraction TRISO-based fuel in light water reactors

For a decade, fully ceramic microencapsulated (FCM) fuel, containing tri-structural isotropic (TRISO) fuel particles in a silicon carbide (SiC) matrix, has been investigated as an accident-tolerant fuel for light water reactors (LWRs). Other examples exist of TRISO-based concepts for LWR fuels with different matrix materials. Previous studies assumed TRISO particle volume packing of approximately 0.44 in SiC (or another) matrix, the highest realistic packing fractions possible with conventional manufacturing. Recent advances in advanced manufacturing have yielded the development and demonstration of a fuel form that consists of conventionally manufactured TRISO particles in a 3D-printed SiC matrix with significantly higher possible TRISO packing fractions (0.5–0.7). This increased uranium loading enhances the viability of using TRISO-based particle fuel forms in LWRs. The viability of high-packing-fraction TRISO-based particle fuel forms in LWRs is assessed from the perspective of fuel cycle length, achievable fuel burnup, reactivity coefficients, and fuel cycle performance. Higher-packing-fraction TRISO-based fuel enables either longer cycle lengths (by ∼25% at a packing fraction of 0.55 relative to 0.44) at a constant enrichment or decreased enrichments (by ∼25% at a packing fraction of 0.55 relative to 0.44) at a constant cycle length. Studies of different fuel kernel types (uranium nitride, uranium oxycarbide, and uranium carbide) yield similar results, although the cycle length of uranium oxycarbide is shorter than for uranium nitride or uranium carbide (due to the lower density of uranium oxide). This work also characterized the production of 14C resulting from neutron absorption in 14N during operation for uranium mononitride fuel kernels; the ratio of 14C/N was 1–2 at. % at discharge. For the fuel cycle evaluation, the activity of spent nuclear fuel and high-level waste at 100 and 100,000 years was lower for high-packing-fraction fuels than for conventional LWR fuel. Environmental impact metrics were similar overall, but higher on the front end of the fuel cycle and lower on the back end of the fuel cycle. Reactivity coefficients of higher-packing-fraction TRISO-based fuel were reasonable compared with those of conventional fuels.

Stress Corrosion Cracking in Austenitic Stainless Steels in Reactor Primary Water–High Purity Oxidizing and in Supercritical Conditions

Abstract Austenitic stainless steels (SS) commonly used in the primary circuit of light water reactors (LWRs) have excellent corrosion resistance in demineralized (DM) water at high temperature (290°C–330°C) and pressure (7.4 MPa for boiling water reactor [BWR], 16 MPa for pressurized water reactor [PWR]). Radiolysis of primary DM water in BWR forms 200–300 ppb of oxidizing species (normal water chemistry [NWC]), which elevates the electrochemical potential of the SS from −300 to −600 mVSHE (mV with standard hydrogen electrode) when dissolved oxygen is 10–20 ppb to +100 to +200 mVSHE. Neutron irradiation of SS further induces metallurgical and microstructural changes, which compromises corrosion resistance. Thus, radiation makes BWR (NWC) environment hostile, causing extensive irradiation-assisted stress corrosion cracking (same as intergranular stress corrosion cracking [IGSCC]) in austenitic SS. IGSCC occurs in both sensitized (grain boundary chromium depletion) and nonsensitized conditions (strain-hardened region in base metal immediately adjacent to the weld fusion zone) and is a generic problem. IGSCC in BWR can be mitigated by hydrogen addition in water (hydrogen water chemistry [HWC]). BWR-HWC has limitations and is ineffective where boiling occurs. IGSCC of SS in PWR is limited because hydrogen addition in primary water suppresses radiolysis though the formation of aggressive environment in low-flow occluded regions can cause IGSCC in SS. Increasing demand for economic power has led to the primary environment in nuclear reactors to become hostile. Conceptual supercritical water reactor (SCWR) will use DM water at 500°C–600°C and 25 MPa, which is extreme for conventional LWR materials. IGSCC in austenitic SS and nickel-based alloys occurs in both oxidizing and reducing SCWR conditions, which can be further exacerbated by radiation. This article reviews how benign primary DM water becomes hostile in LWRs and extreme in SCWR conditions, causing SCC to be a generic problem in austenitic alloys.

The Study of Floating Nuclear Power Plant Reactor Core Neutronic Parameters Using Scale 6.1 Code

KLT-40S nuclear reactor is a small modular floating nuclear power plant made by Russia as a conventional light water reactor (LWR) problem solution nowadays, such as high overnight cost, long commissioning period, and lack of flexibility in supplying a small load of electricity and supplying electricity to isolated areas. With those characteristics, the KLT-40S is suitable to be applied to isolated areas with a small electrical load like archipelagic states such as Indonesia. Based on that reason, Indonesia needs to assess the KLT-40S floating nuclear power plant feasibility study through explorative research. One of those is assessing the reactor core neutronic parameter. In this research, the reactor core modelling is done by using the KENO VI-A and T-6DEPL module in SCALE 6.1 code package. Several variations of reactor operating parameters such as fuel composition and configuration, fuel temperature, moderator temperature, and moderator void fraction had been done in this research. The aim was to get several neutronic parameters to confirm the core feasibility from operational and inherent safety perspectives. Those neutronic parameters are fuel cycle length and reactivity feedback coefficient of fuel temperature, moderator temperature, and moderator void fraction. Based on this research result, the fuel configuration that produces 28 months of cycle length is the fuel base of dispersed UO2 in the silumin matrix with 18,6 % 235U enrichment. Both of the two fuel bases used in this research have inherent safety characteristics, which are shown by the negative value of the reactivity feedback coefficient of fuel temperature, moderator temperature, and moderator void fraction. Dispersed UO2 in the silumin matrix fuel base has better inherent safety characteristics than the UO2 ceramic metal fuel base.

Uncertainty Quantification and Propagation of Multiphysics Simulation of the Pressurized Water Reactor Core

In recent years, the demand to provide best-estimate predictions with confident bounds is increasing for the nuclear reactor performance and safety analysis. The Organisation for Economic Co-operation and Development Nuclear Energy Agency has been developing an international benchmark of the light water reactor (LWR) uncertainty analysis in modeling (UAM) for the examination of uncertainty quantification and propagation methodologies with various modeling and simulation code systems. The objective of the present work is to develop an uncertainty propagation mechanism based on the stochastic sampling method by taking into account the uncertainties of both basic nuclear data and fuel modeling parameters in the simulation of pressurized water reactors (PWRs) that can be incorporated in the conventional LWR simulation approach. More specifically, the Three Mile Island Unit 1–related exercises from the LWR-UAM benchmark have been modeled using the coupled TRACE/PARCS code system in the three-dimensional core representation. The input uncertainties of the neutronics simulation include few-group cross sections and kinetics parameters generated using the Sampler/Polaris sequence of SCALE 6.2.1. Several heat transfer–related variables for the fuel modeling were considered as sources of input uncertainty of the thermal-hydraulics simulations, including the thermal conductivity of fuel and cladding, fuel heat capacity, and the gap conductance. Dakota was used to sample input parameters of the coupled code system and to perform the uncertainty analysis. Two types of simulations were conducted: steady-state calculation at hot full-power condition and transient scenario initiated by the spatially asymmetric rod ejection accident. Quantities of interest for the steady-state calculation, including core multiplication factor and power peaking factors, were calculated with associated uncertainties. For transient calculations, best-estimate plus uncertainty results of the time evolution of core reactivity, core power, and peak fuel temperature were generated and analyzed. The Wilks’ formula was used to determine the necessary sample size to achieve a 95% confidence of the 95% limit of output quantities of interest. Although the uncertainty propagation and quantification method presented in this paper was developed for PWRs, it could be in general applicable to the multiphysics uncertainty quantification of other types of LWR cores.

Effect of hydrogen on corrosion properties of reduced activation ferritic/martensitic steel, F82H

Reduced activation ferritic/martensitic (RAFM) steel, e.g., F82H, is the leading candidate structural material for fusion blanket. Of many blanket concepts, the water-cooled ceramic breeder blanket is an attractive concept because of its compactness and its compatibility with the technologies which using high-temperature water as coolant in conventional light water reactor. For tritium breeding, it is necessary to manage the corrosion of F82H by controlling not chemicals but hydrogen and oxygen in the cooling water. Understanding the effects of oxygen and hydrogen on corrosion property was therefore required for the blanket design. This work aims to investigate the corrosion behavior of F82H in high-temperature water with hydrogen added. Based on the results of the corrosion test, it is speculated that hydrogen addition expected no considerable change of the flow-accelerated corrosion (FAC). In contrast, it was found that the number of cycles to fracture was decreased in the high-temperature water in comparison to the fatigue test at room temperature in air.

Large-scale energy production with thorium in a typical heavy-water reactor using high-grade plutonium as driver fuel

Thorium can be introduced into the energy vector in combination with high-grade plutonium (HG-Pu). Excellent neutron economy of heavy-water moderator allows use of mixed ThO2/HG-PuO2 fuel in heavy-water reactors with high efficiency, leading to the exploitation of large world thorium reserves and extending the availability of the nuclear energy by two orders of magnitude. In the present work, the criticality calculations have been performed with the code MCNPX 3-D geometrical modeling of a typical heavy-water reactor, where the structure of all fuel rods and bundles is represented individually. In the course of time calculations, nuclear transformation and radioactive decay of all actinide elements as well as fission products are considered. Five different fuel compositions have been selected for investigations: (1) 97% thoria (ThO2) + 3% PuO2; (2) 96% ThO2 + 4% PuO2; (3) 95% ThO2 + 5% PuO2; (4) 94% ThO2 + 6% PuO2; and (5) 92% ThO2 + 8% PuO2. The behavior of the criticality k∞ and the burn-up values of the reactor have been pursued by full power operation at 640 MWel (2180 MWth) for approximately 7 years. Time calculations have been conducted with MCNPX and CINDER codes under consideration of all nuclear transformation and radioactive decay processes on the actinide isotopes and fission fragments. As the reactor allows fuel recharging at on-power operation mode, the reactor criticality has been followed down to keff,end = ~1.05. The corresponding burn-up values and operation periods for the investigated modes are (1) 18 GWd/MT and 780 days; (2) 27 GWd/MT and 1200 days; (3) 35 GWd/MT and 1560 days; (4) 44 GWd/MT and 1940 days; and (5) 60 GWd/MT and 2640 days. Among the investigated four modes, 94% ThO2 + 6% PuO2 seems a reasonable choice under consideration of the high price of the HG-Pu as driver fuel. The mixed fuel has the potential of an extensive exploitation of thorium resources. Reactor will run with the same fuel charge for approximately 5 years and allow a fuel burn-up approximately 44 GWd/MT, comparable with conventional light-water reactors (LWRs). Plutonium component of the mixed fuel will become nonprolific after few months of plant operation through the accumulation of even isotopes. Addition of few percent natural uranium to the initial mixed fuel charge will keep the 233U component below 22% and hence at nonprolific level over the entire plant operation period. Replacement of 4% ThO2 with 4% nat-UO2 will practically not change main technical parameters of the reactor.

Monte Carlo analysis of LWR spent fuel transmutation in a fusion-fission hybrid reactor system

Abstract The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.

Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

In a conventional light water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under the assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for Low Enriched Uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields, in the emitted antineutrino spectra, and the detector response function, we show that the difference in core-wide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

FITXS: A fast and flexible burn-up scheme based on the fitting of one-group cross-sections

The closure of the nuclear fuel cycle is currently envisaged with the deployment of Generation IV fast reactors which are able to generate their fuel from fertile 238U or 232Th, and also burn minor actinides arising from legacy wastes and thermal reactors in the nuclear park. The optimization of such fuel cycle strategies requires detailed models, capable of simulating the transition from initial state to equilibrium. Due to the high computational cost of detailed burn-up calculations, most scenario codes use parametrized few group cross-sections to calculate fuel depletion in the reactors. This paper presents a fast and flexible burn-up scheme called FITXS which we applied on the GFR2400 Gas-cooled Fast Reactor and a European Pressurized Reactor MOX fuel assembly. Based on the fitting of one-group cross-sections as functions of the detailed fuel composition, the developed models are able to calculate spent fuel compositions with high accuracy for a wide range of initial compositions in less than one second computational time. In order to demonstrate applicability, the models were integrated into a fuel cycle model containing GFR2400 and European Pressurized Reactors, as well as conventional Light Water Reactors, and the fuel utilization and transmutational properties of the system were analyzed. It was found that the GFR2400 can operate in a closed fuel cycle, with positive breeding gain in the equilibrium. Increased minor actinide feed improves breeding in the GFR2400, but the accumulation of plutonium can be prevented by recycling the excess plutonium as MOX fuel in thermal reactors.

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

Small modular reactors (SMRs) may offer potential benefits relative to large light water reactors, such as enhanced flexibility in deployment and operation. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary neutronics and fuel cycle analyses conducted in support of the US Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies is a “cartridge type” one-batch core with slightly less than 5.0% enrichment.The high-level issues identified and preliminary scoping calculations in this paper are intended to inform decision makers regarding potential fuel cycle impacts of one-batch thermal-spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and a reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for the simplified example SMR analyzed herein are compared with those for a conventional three-batch light water reactor (LWR) in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of per energy generated, mass of depleted uranium disposed of per energy generated, land use per energy generated, and carbon emissions per energy generated.Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores, which mean high leakage (there is a radial and an axial component); (2) a heterogeneous core and extensive use of control rods and burnable poisons; and (3) single-batch cores. But not all SMR designs have all of these traits. The approach used in this study is an example bounding case, and not all SMRs may be impacted to the same extent.

Improving the energy efficiency of NPP

The objective of the present study is the analysis and assessment of potential ways for enhancing the energy efficiency of nuclear power generation.Currently used and promising advanced thermodynamic cycles of nuclear power plants in power generation are examined. Ways for enhancing parameters of working agents in the NPP steam-turbine plants are suggested. Thermodynamic assessment was made of enhancement of thermal efficiency of NPP equipped with fast reactors due to the increase of coolant temperature at the reactor core outlet and, correspondingly, the increase of temperature and pressure of steam generated by steam generator. Analysis of efficiency of utilization of waste low-potential heat at the NPP using heat pumps was performed.Contemporary level of development of machine building industry for power generation, development of high-efficiency high-temperature gas turbines and steam compressors allow addressing the possibility to achieve high conjugated parameters of steam at NPPs equipped with conventional light-water reactors without exceeding the permissible conditions for operation of reactor cores with fuel cladding made of zirconium alloys.Application of heat pumps within the cooling circuit of the main condenser of steam-turbine power plant for the purpose of enhancement of financial indicators is not justifiable. As it was demonstrated by estimation calculations, the use of heat pumps within the main condenser loop is promising only for reducing heat discharged by NPPs in the environment. Utilization using heat pumps of low-potential heat removed by equipment cooling systems is the efficient way to improve NPP efficiency. Non-productive extraction of steam from the thermal cycle of the power unit is reduced in this case which results in the additional power generation as well as reduces heat disposal in the environment.

Utilization of nuclear waste plutonium and thorium mixed fuel in candu reactors

Summary Spent nuclear fuel out of conventional light water reactors contains significant amount of even plutonium isotopes, so called reactor grade plutonium. Excellent neutron economy of Canada deuterium uranium (CANDU) reactors can further burn reactor grade plutonium, which has been used as a booster fissile fuel material in form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates incineration of nuclear waste and the prospects of exploitation of rich world thorium reserves in CANDU reactors. In the present work, the criticality calculations have been performed with 3-D geometrical modeling of a CANDU reactor, where the structure of all fuel rods and bundles is represented individually. In the course of time calculations, nuclear transformation and radioactive decay of all actinide elements as well as fission products are considered. Four different fuel compositions have been selected for investigations: ① 95% thoria (ThO2) + 5% PuO2, ② 90% ThO2 + 10% PuO2, ③ 85% ThO2 + 15% PuO2 and ④ 80% ThO2 + 20% PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the criticality k∞ and the burnup values of the reactor have been pursued by full power operation for ~10 years. Among the investigated four modes, 90% ThO2 + 10% PuO2 seems a reasonable choice. This mixed fuel would continue make possible extensive exploitation of thorium resources with respect to reactor criticality. Reactor will run with the same fuel charge for ~7 years and allow a fuel burnup ~55 GWd/t. Copyright © 2016 John Wiley & Sons, Ltd.

The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe – Part I: Uranium fuel cycle

Disposal of spent nuclear fuel is a major political and public-perception problem for nuclear energy. From a radiological standpoint, the long-lived component of spent nuclear fuel primarily consists of transuranic (TRU) isotopes. Full recycling of TRU isotopes can, in theory, lead to a reduction in repository radiotoxicity to reference levels corresponds to the radiotoxicity of the unburned natural U required to fuel a conventional LWR in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. This strategy forms part of many envisaged ‘sustainable’ nuclear fuel cycles. However, over a limited timeframe, the radiotoxicity of the ‘final’ core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. The importance of low reprocessing losses and minor actinide (MA) recycling is also dependent on the timeframe during which actinides are recycled. In this paper, the fuel cycle code ORION is used to model the recycle of light water reactor (LWR)-produced TRUs in LWRs and sodium-cooled fast reactors (SFRs) over 1–5 generations of reactors, which is sufficient to infer general conclusions for higher numbers of generations. Here, a generation is defined as a fleet of reactors operating for 60 years, before being retired and potentially replaced. Over up to ∼5 generations of full actinide recycle in SFR burners, the final core inventory tends to dominate over reprocessing losses, beyond which the radiotoxicity rapidly becomes sensitive to reprocessing losses. For a single generation of SFRs, there is little or no advantage to recycling MAs. However, for multiple generations, the reduction in repository radiotoxicity is severely limited without MA recycling, and repository radiotoxicity converges on equilibrium after around 3 generations of SFRs. With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance – which appears challenging from an economic and energy security standpoint. TRU recycle in pressurized water reactors (PWRs) with zero net actinide production provides similar performance to low-enriched-uranium (LEU)-fueled LWRs in equilibrium with a fleet of burner SFRs. However, it is not possible to reduce the TRU inventory over multiple generations of PWRs. TRU recycle in break-even SFRs is much less effective from a point of view of reducing spent nuclear fuel radiotoxicity.

A preliminary approach to the ALFRED reactor control strategy

In this paper, a preliminary approach to the definition of a suitable control strategy for the Advanced Lead Fast Reactor European Demonstrator (ALFRED), developed within the European 7th Framework Program, has been undertaken. The Generation IV reactors offer new challenges for what concerns the nuclear power plant control since several constraints both on primary and secondary loops have to be faced, differently from the conventional Light Water Reactors. A simulator of the ALFRED plant has been developed in a previous work (Ponciroli et al., 2014) with the main purpose of studying the system free dynamics and stability features in a control-oriented perspective. Based on the outcomes of these investigations, in the present work, the possibility of adopting decentralized control schemes has been investigated. Accordingly, Single Input Single Output control laws have been applied directly to the selected couples of input–output variables, which have been identified first on the basis of the preliminary plant dynamics analyses, and then confirmed by the indications of the Relative Gain Array method. Afterwards, two different control schemes have been studied depending on the number of available inputs, and then implemented and compared in order to evaluate the effect of each control action on the associated potential control strategy effectiveness. As a last step, the ALFRED control system has been finalized. The regulator design has been set up based on a simultaneous feedforward-feedback scheme incorporating four closed feedback loops. A controlled power reduction and a controlled overpower transient have been simulated in order to assess the performance of the two proposed control schemes. Results show that both the adopted control strategies can assure an efficient control of the thermal power while guaranteeing an effective control of lead and steam temperatures as well. In addition, some non-negligible differences between the two schemes have been observed and discussed in the simulation results of control and controlled variables.

Start-up transient test simulation with and without void-reactivity feedback for a two-phase natural circulation reactor

The new small-scaled light water reactor design, known as integral modular water reactor (IMR), has an advantage compared to conventional light water reactor in terms of costs by adopting innovative technologies. However, this reactor may be susceptible to flow instabilities due to two-phase natural circulation inside the reactor pressure vessel (RPV). Flow instabilities may be amplified due to strong interaction between the flow and core power through the void-reactivity feedback mechanism. During the start-up of the IMR, system pressure is low. At low pressure, the density ratio can be quite high, which leads to large variation in void fraction due to change in flow quality. In the IMR design, the long riser and large volume of water can lead to thermal non-equilibrium between the phases due to the significant variation of the saturation temperature along flow direction. This can result in flow oscillations at certain operating conditions.In order to understand and identify the instability phenomena during the start-up of the reactor, a scaled experimental facility is designed based on the sound scaling approach. The scaling laws are used to obtain design parameters to maintain the similarities between prototype and experimental facility. Modified power scaling is used for transient tests. Four heater rods are used to simulate the chaotic flashing phenomena. In order to investigate the instability during the start-up procedure, start-up transient tests are performed with and without void-reactivity feedback for different core heat-up rate. These experimental results show that IMR is susceptible to instability during the start-up procedure at very low pressure. These oscillations are condensation induced oscillations and occur due to high subcooling in chimney section. It is investigated that, since amplitude of power oscillation is low, void-reactivity feedback effect does not affect the stability of the system for this type of given power transient.

TERRAPOWER, LLC TRAVELING WAVE REACTOR DEVELOPMENT PROGRAM OVERVIEW

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ∼30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

Evaluation of neutron economical effect of new cladding materials in light water reactors

The present paper aims to contribute from a neutronic aspect to activities for new cladding material development for light water reactors (LWRs) that can reduce the risk of hydrogen gas explosion. Iron (Fe), nickel (Ni), titanium (Ti), niobium (Nb) and vanadium (V) are selected as possible component elements to cover a variety of new cladding materials for LWRs. The effect of larger thermal absorption cross sections of these elements than those of zirconium (Zr), together with those of silicon carbide (SiC), on the neutron economy in LWRs is evaluated by performing pin cell burnup calculations for a conventional pressurized water reactor (PWR), a low-moderation high-burnup LWR (LM-LWR) and a high-moderation high-burnup LWR (HM-LWR). As can be anticipated from the thermal cross sections, SiC has excellent neutron economy. The materials other than SiC largely decrease discharge burnup for all three types of LWRs in comparison with Zircaloy-4. Among such elements of larger thermal absorption cross section, Nb has neutron economical advantage over the other materials except SiC in softer neutron spectrum reactors such as HM-LWR in which the atomic number ratio of hydrogen to heavy metal is 6. In conventional LWRs, stainless steel of low Ni contents has the neutron economic advantage as well as Nb for cladding material. The results of the calculations are summarized for the purpose to provide reference data for new cladding material development studies, in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding.

Thorium fuel production and results from beginning of life irradiation

An evolutionary, rather than a revolutionary approach to thorium fueled reactors is proposed with an introduction of thorium as the fertile component in mixed oxide fuel for conventional light water reactors. The utility of thorium as a component in today's light water reactors offers improved accident tolerance due to the superior material properties of thorium fuels over conventional uranium fuels. A few notable advantages include better thermal conductivity, higher melting point, higher oxide stability and superior spent fuel characteristics. Consequently, Thor Energy along with a consortium of industrial partners has established a fuel production and irradiation program aimed to license thorium fuels for use in today's light water reactors. Due to the morphology and physical properties of calcined thorium oxide powder, pellet fabrication used for standard uranium oxide fuels must be altered slightly for thorium bearing fuels to yield a product with acceptable theoretical densities, microstructure, and material integrity. At beginning of irradiation life fuel temperature data demonstrates improved fuel characteristics over standard uranium oxide fuel. Fuel centerline operating temperatures are 30–40 K less with a thorium mixed oxide fuel as compared to standard uranium fuel.