Internal Blanket Research Articles

不同保温覆盖措施对日光温室温度环境的影响

Chinese solar greenhouses (CSGs) are the important agricultural building facility with highly efficient and sustainable utilization of solar energy. In order to improve the thermal insulation performance and reduce the heat loss, different insulation cover measures were applied to Chinese solar greenhouses. In this study, a solar greenhouse with an internal insulation blanket (SG1), a greenhouse with an internal blanket and an internal plastic film (SG2) and a greenhouse with an external blanket, an internal blanket and an internal plastic film (SG3) were chosen as experimental objects to analyze their indoor temperature environment. The results indicated that the differences of solar radiation interior among three greenhouses were similar, which was not the main reason that caused the differences of the temperature environment in the three greenhouses. During the tests, the nighttime average temperature was increased by 2.1-3°C by adding the cover materials, and the duration of nighttime temperature greater than 8°C was increased by 33.5%-38.2%. In addition, the cost is increased by adding the insulation cover materials, but it will be beneficial for greenhouse production to reduce additional energy consumption

A safety enhanced sodium-cooled MOX fueled fast reactor core concept

ABSTRACT We have developed a 750MWe sodium-cooled MOX fueled fast reactor core with enhanced safety against typical ATWS. Design goals were set as follows: 1) negative void reactivity for the loss of flow (ULOF); and 2) burnup reactivity less than 1 $ for the transient over power (UTOP). To achieve the goals the following were considered: 1) adoption of the axially heterogeneous core with a sodium plenum and a gas expansion module to reduce void reactivity; and 2) addition of minor actinides (MAs) to the internal blanket to reduce burnup reactivity. Configurations of the core were optimized as follows. First, void reactivity was reduced by a parameter survey for the inner core height, B-10 content of the upper shield, and standby position of the backup control rods. Second, burnup reactivity was reduced by a parameter survey for the MA content in the internal blanket, etc. We designed the safety enhanced fast reactor core that did not undergo sodium boiling if the ULOF occurred and we maintained fuel integrity for the UTOP by the transient analysis. By installing a self-actuated shutdown system in the backup control rods, the improved operational reliability of this shutdown system and the improved safety can be expected.

Development of a core design capability for innovative boiling water reactor designs for burning transuranic isotopes using WIMS/PANTHER

The resource-renewable boiling water reactor (RBWR) is an innovative BWR design proposed by Hitachi that has the capability to recycle actinides fully, either in a self-sustaining fuel cycle or for the purposes of reducing the environmental impact of long-lived transuranic isotopes (TRUs). Compared to a conventional BWR, a harder neutron spectrum is achieved through utilizing a tight pitch, triangular lattice and increasing the core average void fraction. An axial seed-blanket structure is utilized, with an internal blanket placed between two regions with high TRU loading. In this paper, two TRU-burning configurations of RBWR are considered: the RBWR-TB2, which contains hexagonal assemblies, and the RBWR-BF (‘backfit’), which has the goal of increasing the compatibility with existing or planned ABWRs through utilizing the same assembly pitch and shape. A reactor physics modelling capability is developed for these two diverse configurations of RBWR using the UK ANSWERS reactor physics code WIMS and the PANTHER whole core analysis code. The developed models are then applied to the analysis of these RBWR concepts. For the RBWR-TB2, results are consistent with previous work in predicting reactivity coefficients and core power distribution. In particular, the void coefficient is calculated to be negative. WIMS-PANTHER is then applied to the novel case of the RBWR-BF, for which a basic conceptual design is modelled. The void coefficient for the RBWR-BF is calculated to be negative. The developed methodology is relatively fast running and hence is well suited to performing future design studies, in particular for the relatively new concept of the RBWR-BF.

Three Confinement Systems—Spherical Tokamak, Standard Tokamak, and Stellarator: A Comparison of Key Component Cost Elements

Since the 1950s “next-step” fusion devices and power plant studies have been developed for a number of magnetic confinement systems but an open question remains: can a magnetic fusion device be simplified to the point where it will be cost competitive and operate with high availability? Concept designs based on the ARIES advanced tokamak, spherical tokamak (ST), and the quasi-axisymmetric (QAS) stellarator option have progressed in recent years through a series of Princeton Plasma Physics Laboratory (PPPL) studies with an underlying intent to improve the engineering feasibility of each, giving special attention to concepts that simplify the device configuration and improve maintenance features. For the ST option, design details centered on a 3-m Fusion Nuclear Science Facility that evolved to incorporate vertical maintenance, high temperature superconductor magnets, a small inboard duel coolant lead lithium blanket, and a liquid metal divertor. In collaboration with the Korean fusion demonstration reactor (K-DEMO) and CFETR concept study teams the tokamak design has evolved to increase plasma component access within a vertical maintenance approach using enlarged toroidal field coils incorporating a low- and high-field Nb3Sn winding pack that provide a peak field of 16 T. A recent PPPL stellarator study focused on simplifying the stellarator winding topology to improve access to in-vessel components; combining coil optimization with winding surfaces that incorporated geometry constraints specified by engineering. This paper centered on a 1000-MW power plant design with a tokamak like vertical maintenance scheme that allows access to remove large internal blanket sectors. Results of three confinement studies (PPPL developed ST, K-DEMO, and QAS stellarator) will be presented to highlight concepts that simplify each device configuration and improved their maintenance features. Scaling each option to a common 1000-MW net electric power plant mission allows comparisons to be made of key cost elements such as the size of major core components, sizing of the reactor hall, or external facilities needed to handle and store activated in-vessel components.

Core Design and Analysis of Axially Heterogeneous Boiling Water Reactor for Burning Transuranium Elements

The resource-renewable boiling water reactor (RBWR) is an innovative boiling water reactor that has the capability to breed or to burn transuranium elements (TRUs). Core characteristics of the RBWR of the TRU burner type were evaluated by two different core analysis methods. The RBWR core features an axially heterogeneous configuration, which consists of an internal blanket region between two seed regions, to achieve the TRU multi-recycling capability while maintaining a negative void reactivity coefficient. Axial power distribution of the TRU burner core tends to be more heterogeneous because the isotopic composition ratio of fertile TRUs to fissile TRUs becomes larger in the TRU burner–type core than in the breeder-type core and the seed regions need to be axially shorter than that of the breeder-type core. Thus core analysis of the TRU burner–type core is more challenging. A conventional diffusion calculation using nuclear constants prepared by two-dimensional lattice calculations was performed by Hitachi, while the calculation using nuclear constants prepared by three-dimensional calculations and axial discontinuity factors was performed by the University of Michigan to provide a more sophisticated treatment of the axial heterogeneity. Both calculations predicted similar axial power distributions except in the region near the boundary between fuel and plenum. Both calculations also predicted negative void reactivity coefficients throughout the operating cycle. Safety analysis was performed by Massachusetts Institute of Technology for the all-pump trip accident, which was identified as the limiting accident for the RBWR design. The analysis showed the peak cladding temperature remains below the safety limit. Detailed fuel cycle analysis by University of California, Berkeley, showed that per electrical power generated, the RBWR is capable of incinerating TRUs at about twice the rate at which they are produced in typical pressurized water reactors.

Trade-Off Studies for the Fuel-Self-Sustaining RBWR-Th Core

This study searches for the optimal fuel assembly design for the RBWR-Th core, which is a reduced-moderation boiling water reactor that is fuel-self-sustaining. Except for the initial fuel loading, it is charged with only fertile fuel and discharges only fission products, recycling all actinides. The RBWR-Th is a variant of the RBWR-AC core proposed by Hitachi, which arranges its fuel in a hexagonal tight lattice, has a high outlet void fraction, axially segregates seed and blanket regions, and fits within the advanced boiling water reactor (ABWR) pressure vessel. The RBWR-Th shares these characteristics but replaces depleted uranium (DU) with thoria as the primary fertile fuel, eliminates the internal blanket while elongating the seed region, and eliminates absorbers from the axial reflectors.The sensitivity of important RBWR-Th core performance parameters to change in each one of a dozen design variables was established. These sensitivities provide useful insight and guidance to search for the optimal core design. The design variables of the sensitivity studies include the length of the seed and blanket zones, fuel rod diameter, lattice pitch, number of pins per assembly, concentration distribution of the recycled transfertile (transuranium + transthorium) isotopes in the seed, amount of DU in the seed makeup, coolant mass flow rate, and simulated depletion cycle length. The performance of the RBWR-Th core was found to be highly sensitive to the pitch-to-diameter ratio and to modeling assumptions. Using the conservative modeling assumptions, it was not possible to get the full ABWR power level without exceeding the pressure drop constraint.

Self-sustaining Thorium-fueled Reduced Moderation BWR Feasibility Study

Abstract This study assesses the feasibility of a thorium-based fuel-self-sustaining Reduced moderation BWR (RBWR-Th) core. This core features a hexagonal tight-lattice fuel, high exit coolant quality, axial segregation of seed and blanket regions, and compatibility with the ABWR reactor vessel inherited from the RBWR-AC proposed by Hitachi. The RBWR-Th differs from the RBWR-AC by eliminating the internal blanket, eliminating absorbers from the axial reflectors, replacing depleted urania with thoria as the primary fertile fuel and elongating the fissile region. When the Hitachi assumptions and correlations for thermal-hydraulic analysis are used, the RBWR-Th obtains an average core discharge burnup of 61 GWd/t versus 45 GWd/t for the RBWR-AC. However, when more conservative thermal-hydraulic correlations are used, the average discharge burnup drops to 25 GWd/t. The RBWR-Th maintains negative coolant void coefficients of reactivity throughout the cycle – between -155 and -131 pcm/% – but currently does not have sufficient margin for cold shutdown. Future studies will address this shutdown margin deficit and will account for coolant pressure drop constraint.

Self-Sustaining Thorium Boiling Water Reactors

A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR) proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio. In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio. In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.

Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40–50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mm×2 , internal blanket of 150 mm and axial blanket of 400 mm×2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in ( seed+ internal blanket + outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mm×2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature coefficient is negative for both of cases.

Conceptual designing of reduced-moderation water reactor with heavy water coolant

The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06–1.11.

A Batch Extraction Method for Pulsed Multi-Flux Activation Calculations

A new practical method has been developed for calculating neutron-activation inventories of target material in inertial fusion energy (IFE) reactors such as HYLIFE-II. It accounts for irradiation both in the target and in the internal blanket and for material circulation in and out of the primary loop. The continuous removal of target material in the real system is approximated by a batch extraction (BE). A single target is followed through its lifetime in the reactor using “transition matrices” for activation and decay which are generated by the ACAB code package. The inventory of all the isotopes of interest accumulating in the reactor is obtained by superimposing the contribution of single targets. The new BE model simulates, within minutes, the evolution of more than 150 isotopes over the 30-year reactor lifetime, explicitly accounting for the millions of neutron pulses experienced by a single target and summing the inventories of all the targets.

Measurement of Flux Tilt and Eigenvalue Separation in Axially Decoupled Core

A study on the nuclear characteristics of an axially decoupled core was performed through a series of experiments using the solid moderator core in the Kyoto University Critical Assembly (KUCA). In the axially decoupled core which has an internal blanket between the upper and lower cores, the measured differential reactivity worth of a control rod moving axially was apparently asymmetric between the two core regions. Since it was considered that this phenomenon was attributed to the neutron flux tilt caused by the axial movement of control rods during the measurement of differential worth, the axial reaction rate distributions were measured by gold wires with and without a cadmium tube. It was found that the reaction rate distribution was remarkably tilted by changing the pattern of control rods inserted in the core to maintain the critical state, and that the flux distribution was tilted without energy dependency. The value of first-order eigenvalue separation was obtained using the relation between the neutron flux tilt and the first-mode eigenvalue separation based on the Explicit Higher-order Perturbation (EHP) method. This value agreed with the calculated result by the SRAC system.

Feasibility Studies on Plutonium and Minor Actinide Burning in Fast Reactors

Systematic parameter studies were implemented to investigate the basic characteristics [plutonium and minor actinide (MA)-burning rate, burnup reactivity loss, Doppler coefficient, sodium void reactivity, maximum linear heat rate, etc.] of plutonium and MA-burning fast reactors and also to clarify the feasibility of such plutonium and MA burner fast reactors. Highly enriched mixed-oxide (MOX) fuels and plutonium fuels without uranium were consideredfor plutonium-burning enhancement. It was found that plutonium consumption rates essentially depend on plutonium enrichment. Both burnup reactivity loss and Doppler coefficient are important criteria for highly enriched MOX fuel cores. Cores without uranium were found to consume the plutonium at a very large burnup rate close to the theoretically maximum value of 110 to 120 kg/TW · h(electric). The introduction of UO2 in an internal blanket is effective in enhancing the Doppler coefficient; it causes a minor increase in the sodium void reactivity in nonuranium cores. The MA transmutation in a fast reactor core has no serious drawbacks in terms of core performance, provided that the homogeneous loading method can be employed with a small fraction of MA fuel (∼5 wt%). Fast reactors have a strong potential for burning plutonium and MA effectively.

Parametric Study on Fast Reactors with Low Sodium Void Reactivity by the Use of Zirconium Hydride Layer in Internal Blanket

A study on the sodium void reactivity reduction of 1,000MWe class fast reactors was performed. The target of this study is to materialize a core concept with a zirconium hydride layer in the internal blanket which satisfies the design criteria; that the possibility of the prompt critical by sodium voiding can be excluded at any time during reactor operation; and that the burnup reactivity is acceptably low, which is less than 2.5%dk/kk′. A series of parametric surveys were performed on the radial size of the internal blanket and the width of the zirconium hydride layer. The width of the layer was set to 3.0cm as the optimized specification which satisfies all the design criteria regardless of the fuel materials; oxide, nitride and metal.Transient behaviors of the cores were investigated during the unprotected loss of flow accident (LOF). The analyses showed that the oxide core was able to avoid sodium boiling. It was also observed that the cores with nitride and metal fuels had larger tolerance against the unprotected LOF: The outlet temperature were lower by 39 and 60°C compared with the oxide core.

Parametric Study on Fast Reactors with Low Sodium Void Reactivity by the Use of Zirconium Hydride Layer in Internal Blanket.

A study on the sodium void reactivity reduction of 1,000MWe class fast reactors was performed. The target of this study is to materialize a core concept with a zirconium hydride layer in the internal blanket which satisfies the design criteria; that the possibility of the prompt critical by sodium voiding can be excluded at any time during reactor operation; and that the burnup reactivity is acceptably low, which is less than 2.5%dk/kk′. A series of parametric surveys were performed on the radial size of the internal blanket and the width of the zirconium hydride layer. The width of the layer was set to 3.0cm as the optimized specification which satisfies all the design criteria regardless of the fuel materials; oxide, nitride and metal. Transient behaviors of the cores were investigated during the unprotected loss of flow accident (LOF). The analyses showed that the oxide core was able to avoid sodium boiling. It was also observed that the cores with nitride and metal fuels had larger tolerance against t...

Measurement of Flux Tilt and Eigenvalue Separation in Axially Decoupled Core.

A study on the nuclear characteristics of an axially decoupled core was performed through a series of experiments using the solid moderator core in the Kyoto University Critical Assembly (KUCA). In the axially decoupled core which has an internal blanket between the upper and lower cores, the measured differential reactivity worth of a control rod moving axially was apparently asymmetric between the two core regions. Since it was considered that this phenomenon was attributed to the neutron flux tilt caused by the axial movement of control rods during the measurement of differential worth, the axial reaction rate distributions were measured by gold wires with and without a cadmium tube. It was found that the reaction rate distribution was remarkably tilted by changing the pattern of control rods inserted in the core to maintain the critical state, and that the flux distribution was tilted without energy dependency. The value of first-order eigenvalue separation was obtained using the relation between the neutron flux tilt and the first-mode eigenvalue separation based on the Explicit Higher-order Perturbation (EHP) method. This value agreed with the calculated result by the SRAC system.

Disruption problematics in segmented-blanket concepts

In tokamaks, the hostile operating environment originated by plasma disruption events requires that the first-wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence, there is a need to improve the safety features of the segmented-blanket design concepts in order to satisfy the disruption problematics. The present paper describes recent investigations on internal blanket reinforcement systems needed in order to improve the first-wall/blanket/shield structural design for next-step and commercial fusion reactors. Particularly in the context of SEAFP and ITER activities, representative 3-D CAD models of the inboard and outboard blanket regions and the related magnetomechanical simulations are illustrated.

Axial distribution of cesium in heterogeneous FBR fuel pins

The axial migration behavior of cesium was investigated in two kinds of axially heterogeneous (U, Pu)O 2 fuel pins with different configurations of internal blanket position irradiated to $ ̃ 70 GWd/t. The axial distribution of cesium was measured nondestructively by a γ-ray intensity scanning and destructively by an electron probe microanalyzer. A reasonably good agreement between the experimental and analytical results was obtained by applying a model based on a two-stage process proposed by Adamson. The radial distribution could be explained by assuming a rapid migration down temperature gradient within the solid fuel. The evaporation-condensation process not only across the pellet-pellet interfaces, but also across cracks played an important role in the axial migration.