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1.
《Annals of Nuclear Energy》1999,26(17):1517-1535
The sensitivity of various safety parameters, affecting the reactivity insertion limits imposed by clad melting temperature for a typical pool type research reactor, have been investigated in this work. The analysis was done for low enriched uranium (LEU) core with scram disabled conditions. The temperature coefficients of fuel and coolant, void/density coefficient and βeff were individually varied and the reactor behavior for different ramp reactivity transients was studied. In this work ramp reactivity insertions from 1.6 to 2 $/0.5 s were selected and peak power, maximum fuel, clad and coolant temperatures were determined. Results show that peak power decreases with an increase in the Doppler coefficient of reactivity. However, it rises with an increase in the reactivity insertion. Core remains insensitive to the coolant temperature coefficient of reactivity for ramps in the range of 1.6–1.9/0.5 s. Peak power decreases with an increase in the void coefficient of reactivity (0.1 $/%void to 0.8 $/%void). With a decrease in the void coefficient of reactivity, the maximum fuel and clad temperatures show a non-linear rise. Power and temperature peaks in the transient are sensitive to the values of βeff. Finally, it can be concluded that LEU is a safe core due to its smaller βeff, larger Doppler coefficient and void coefficient of reactivity. It is inferred through this work that reactivity insertion limits of LEU core are quite insensitive to βeff, the Doppler coefficient and the coolant temperature coefficient of reactivity. They are highly sensitive to the change of the void coefficient of reactivity in the core. 相似文献
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3.
《Journal of Nuclear Science and Technology》2013,50(11):618-632
Abstract The behavior of neutrons in a highly heterogeneous unit cell consisting of D2O moderator, H2O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different 235U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D2O level, as well as various lattice parameters A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H2O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fuel region produced by the H2O coolant. The results of cell calculations obtained by means of the METHUSELAH-II code showed generally good agreement with experimentally determined data, for both critical D2O levels and lattice parameters, in the case of coolant-filled lattices (0, 30 and 70% void fractions). For lattices devoid of coolant (100% void fraction), however, discrepancies in lattice parameters—particularly in p 28—produced corresponding deviations in core reactivity amounting to 1% in excess of those incurred with other void fractions. 相似文献
4.
The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D2O-moderated, and H2O-, D2O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) underpredicted keff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 overpredicted keff and underpredicted the CVR bias relative to MCNP5 by 100–200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately. 相似文献
5.
We have studied the efficiency of spallation neutron sources for different combinations of coolant and fuel in 80 MWth, sub-critical, cores. It has been found that the proton source efficiency, ψ∗, is reduced by 10% when switching coolant from helium to lead–bismuth eutectic. Substituting MOX fuel with an americium based fuel, results in another 10% reduction of ψ∗. The relatively high source efficiencies found for prototype accelerator-driven systems, using standard MOX fuel and helium coolant, may thus be difficult to achieve in future systems dedicated to the transmutation of higher actinides. Our results are in agreement with previous investigations of the dependence of the source efficiency on the selection of coolant. 相似文献
6.
Chungho Cho Yonghee Kim Tae Yung Song Yong-Bum Lee 《Nuclear Engineering and Design》2008,238(1):90-101
A spallation target system is a key component to be developed for an accelerator driven system (ADS). It is known that a 15–25 MW spallation target is required for a practical 1000 MWth ADS. The design of a 20 MW spallation target is very challenging because more than 60% of the beam power is deposited as heat in a small volume of the target system. In the present work, a numerical design study was performed to obtain the optimal design parameters for a 20 MW spallation target for a 1000 MWth ADS. A dual injection tube was proposed for a reduction of the lead–bismuth eutectic (LBE) flow rate at the target channel. The results of the present study show that a 30 cm wide proton beam with a uniform beam distribution should be adopted for a spallation target of a 20 MW power. When the dual LBE injection tube is employed, the LBE flow rate could be reduced by a factor of 7 without reducing the allowable beam current. 相似文献
7.
《Annals of Nuclear Energy》2005,32(1):13-28
Within the European Fifth Framework Program, Preliminary Design Studies of an Experimantal Accelerator Driven System (PDS-XADS) being supported by the European Commission are focussed on options employing molten Lead–Bismuth Eutectic (LBE) and helium gas coolants. Two of the options employ 80 MWth subcritical cores which are driven by a 600 MeV proton beam with a maximum current of 6 mA, the proton beam impinging either on a window or a windowless LBE target near the core center. By assuming, for example, one-batch operation, the fuel discharge burnup being consistently computed with the deterministic code ERANOS (Version 2.0) is ∼20 MWd/kg for the gas-cooled XADS and ∼25 MWd/kg for the LBE XADS. The larger source importance of the gas-cooled XADS ensures that these two values are relatively close in spite of the more negative reactivity level of the gas-cooled XADS. The gas-cooled XADS exhibits a much larger transport effect reflecting the strong anisotropy of scattering in the low density regions. However, the sensitivity to the characteristics of the external neutron source being precalculated with MCNPX and then used in the burnup calculations is quite small in both cases. 相似文献
8.
Pavel Hejzlar Neil E. Todreas Anna Nikiforova Michael J. Driscoll 《Nuclear Engineering and Design》2009,239(12):2672-2691
Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl2 (LSFR), sodium (SFR), and supercritical CO2 (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO2 power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO2 reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR. 相似文献
9.
Naoto Aizawa Tomohiko Iwasaki Yasuaki Watanabe Tatsuro Takani 《Journal of Nuclear Science and Technology》2016,53(2):240-249
A reactivity control method for accelerator-driven system (ADS) is studied for its ability to reduce both the maximum beam current and the load on the beam window. A burnable position (BP) assembly (with gadolinium and zirconium hydride pins) is applied to the ADS core for reactivity control, and various BP assembly optimizations (such as pin arrangement, BP assembly loading position, and BP composition) are performed to minimize the burnup reactivity swing. These optimizations lead to a decreased burnup reactivity swing that is as much as 82.5% less than the swing of non-BP-loaded core; furthermore, the maximum beam current is 12.5 mA. The reactor characteristics of the optimized BP-loaded ADS core are also analyzed to investigate how the introduction of BP assembly influences the system. Safety parameters (such as the Doppler coefficient and the coolant void reactivity) worsen with the introduction of BP assembly, and the total minor actinide transmutation amount decreases 30–40 kg because of the moderated neutrons and changed fuel composition. 相似文献
10.
An accelerator-driven system (ADS) combined with a subcritical molten salt reactor (MSR) is a type of hybrid reactor originally designed to use Th/U (or U/Pu ) fuel cycles. In most accelerator-driven molten salt reactor (AD-MSR) concepts, the salt material is also used as a target for inducing spallation neutrons. Although a neutron source is an important component in the design of ADS, only a few studies have addressed the effects of the neutron spallation source in the AD-MSR. Incidentally, there is no quantitative study on how much the beam power can be reduced by installing a spallation target in a sodium chloride-based fast reactor. We studied the proton and the neutron source efficiencies of an AD-MSR with chloride fuels by considering an Lead Bismuth Eutectic (LBE) spallation target. This LBE target is found to increase the proton source efficiency significantly. The required beam power for an AD-MSR can be reduced by 33 % and 16 % for NaCl-Th/233U and NaCl-U/Pu fuels, respectively, relative to the AD-MSR without the LBE spallation target by keeping the same keff. The energy gain can be increased up to 1.5 times and 1.2 times for NaCl-Th/233U and NaCl-U/Pu fuels, respectively. Thus, incorporating a spallation target module in an AD-MSR can significantly reduce the burden on the accelerator. 相似文献
11.
Subcritical reactors, also called Accelerator Driven Systems (ADS), are specifically studied for their capacity in transmuting Minor Actinides (MA). Nuclear fuel cycle scenarios involving MA transmutation in ADS are widely researched. The nuclear fuel cycle simulation tool code CLASS (Core Library for Advanced Scenarios Simulations) is dedicated to the inventory evolution calculation induced by a complex nuclear fleet. For managing reactors, the code CLASS includes physic models. Loading models aim to provide the fuel composition at beginning of cycle according to the stocks isotopic composition and the reactors requirements. A cross section predictor aims to provide mean cross sections needed for solving Bateman equations. Physic models are built from reactors calculation set ahead of the scenario calculation. An ADS standard composition at BOC is a mixture of plutonium and MA oxide. The high number of fissile isotopes present in the subcritical core leads to an issue for building an ADS fuel loading model. A high number of isotopic vector at BOC is needed to get an exhaustive simulation set. Also, ADS initial reactivity is adjusted with an inert matrix which induces an additional degree of freedom. The building of an ADS fuel loading model for CLASS requires two steps. For any heavy nuclide composition at beginning of cycle, the core reactivity must be imposed at a subcritical level. Also, the reactivity coefficient evolution should be maintained during the irradiation. In this work, the MgO volume fraction is adjusted to reach the first requirement. The methodology based on a set of reactor simulations and neural network utilization to predict the MgO volume fraction needed to reach a wanted keff for any initial composition is presented. Also, a complete neutronic study is done that highlight the effect on MgO on neutronic parameters. Reactor simulations are done with the transport code MCNP6 (Monte Carlo N particle transport code). The ADS geometry is based on the EFIT (European Facility for Industrial-Scale Transmutation) concept. The simulation set is composed of more than 8000 randomized runs from which a neural network has been built. The resulting MgO prediction method allows reaching a keff at 0.96 and the distribution standard deviation is around 200 pcm. 相似文献
12.
Masao Yamanaka Cheol Ho Pyeon Song Hyun Kim Hiroshi Shiga Yasunori Kitamura Tsuyoshi Misawa 《Journal of Nuclear Science and Technology》2017,54(3):293-300
In the accelerator-driven system (ADS), the effective delayed neutron fraction (βeff) is a requisite for converting the subcriticality from dollar units to pcm units. To evaluate the accurate βeff value in ADS, measurement of βeff complements its calculation methodology and the nuclear data on delayed neutrons. Subcriticality measurements are carried out by the pulsed neutron source method in the Kyoto University Critical Assembly, and the neutron noise analyses are conducted by the Rossi-α method with the pulsed shape of the spallation neutron source. The value of βeff is deduced with the combined use of measured subcriticality in dollar units and correction parameters by MCNP6.1 together with JENDL-4.0 and JENDL/HE-2007. A comparison between the calculated and the measured βeff represents the acceptable accuracy within the subcriticality range of around keff = 0.93 in the ADS operations. Here, the applicability of the measurement methodology based on the Rossi-α method is demonstrated by varying the subcriticality. 相似文献
13.
Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems 总被引:5,自引:0,他引:5
The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. keff 0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. keff0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early design phase of any ADS systems in order to assure a benign transient response of the particular ADS design under investigation to typical plant transient initiators. 相似文献
14.
《Journal of Nuclear Science and Technology》2013,50(9):834-845
Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ?2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25–40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient. 相似文献
15.
Since many years, the Forschungszentrum Karlsruhe has been working on the research and development of an accelerator driven sub-critical system (ADS) cooled by lead–bismuth eutectic (LBE). Although various numerical tools for thermal hydraulics have been established at the Forschungszentrum Karlsruhe, reliably validated physical models related to turbulent heat transfer in LBE flows are still missing. Especially, physical models on heat transfer and turbulent Prandtl number have to be re-evaluated in LBE conditions to improve the reliability of numerical tools. In the present paper, review and assessment of the existing physical models are made. Computational fluid dynamics (CFD) analysis is carried out for circular tube geometries. Based on the assessment of the existing models and the CFD results achieved, recommendations are made on correlations of heat transfer and turbulent Prandtl number for numerical applications to LBE flows. 相似文献
16.
有效缓发中子份额(βeff)、平均中子代时间(Λ)和反应性反馈系数(α)是核反应堆动力学中至关重要的参数。本文采用蒙特卡罗方法计算了加速器驱动的次临界系统(ADS)堆芯的动力学参数,并分析了次锕系核素(MA)装载量对这些参数的影响。通过在燃料中添加不同含量的MA,来研究其对ADS堆芯动力学参数的影响。结果表明,当MA在燃料中的含量从0%增加到5%时,βeff和Λ的值分别降低了18%和31%,多普勒反馈系数平均值α-D由-0.56 pcm/K变化到-0.36 pcm/K,冷却剂反馈系数平均值α-C由-2.11 pcm/K变化到-1.73 pcm/K。 相似文献
17.
《Journal of Nuclear Science and Technology》2013,50(3):483-490
For a dedicated transmutation system, Japan Atomic Energy Agency (JAEA) has been proceeding with the research and development on an accelerator-driven subcritical system (ADS). The ADS proposed by JAEA is a lead-bismuth eutectic (LBE) cooled fast subcritical core with 800 MWth. JAEA has started a comprehensive research and development (R&D) program since the fiscal year of 2002 to acquire knowledge and elemental technology that are necessary for the validation of engineering feasibility of the ADS. In this paper, the outline and the results in the first three-year stage of the program are reported. Items of R&D were concentrated on three technical areas peculiar to the ADS: (1) a superconducting linear accelerator (SC-LINAC), (2) the LBE as spallation target and core coolant, and (3) a subcritical core design and reactor physics of the ADS. For R&D on the accelerator, a prototype cryomodule was built and its good performance in electric field was examined. For R&D on the LBE, various technical data for material corrosion, thermal-hydraulics and radioactive impurity were obtained by loop tests and reactor irradiation. For R&D on the subcritical core, engineering feasibility for the LBE cooled tank-type ADS was discussed using thermal-hydraulic and structural analysis not only in normal operation but also in transient situations. Reactor physics experiments for subcritical monitoring and physics parameters of the ADS were also performed at critical assemblies. 相似文献
18.
This research is focused on using Thorium-Plutonium MOX fuel in the inner fuel pins of the CANDU fuel bundles for plutonium incineration and reduction of uranium demand and to reduce coolant void reactivity. The delayed neutron fraction and the power distribution amongst the fuel elements of the fuel bundle have been considered as main safety parameters.The 700 MWe Advanced CANDU Reactor (ACR-700) was selected as a case study. The inner eight UO2 fuel pins of the ACR-700 fuel bundle are replaced by Thorium-Plutonium MOX fuel pins in the proposed design with 3% reactor grade PuO2. This amount represents 23.4 w/o of the fuel in the bundle. The outer two fuel rings (35 pins) enrichment is reduced from 2.1 w/o U-235 to 2 w/o U-235. The simulation using MCNP6 showed that about 27% reduction of uranium demand can be achieved. The proposed fuel bundle eliminate the use of burnable poisons in the central pin that was used for negative coolant void reactivity and more reduction in the coolant void reactivity was achieved (about 3.5 mk less than the reference fuel bundle). The power distribution throughout the fuel bundle is more flat in the proposed fuel bundle. Use of this fuel bundle reduces the delayed neutron fraction from 540 pcm in the reference case to 480 pcm in the proposed case. 相似文献
19.
Coolant void reactivity (CVR) is an important factor in reactor accident analysis. Here we study the adjustments of CVR at beginning of burnup cycle (BOC) and keff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice using the optimization and adjoint sensitivity techniques. The sensitivity coefficients are evaluated using the perturbation theory based on the integral neutron transport equations. The neutron and flux importance transport solutions are obtained by the method of cyclic characteristics (MOCC). Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR-BOC (CBCVR-BOC). To approximate the EOC sensitivity coefficient, we perform constant-power burnup/depletion calculations using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Our aim is to achieve a desired negative CVR-BOC of −2 mk and keff-EOC of 0.900 for the first two cases, and a CBCVR-BOC of −2 mk and keff-EOC of 0.900 for the last case. Sensitivity analyses of CVR and eigenvalue are also included in our study. 相似文献
20.
加速器驱动的次临界系统(ADS)是未来最有可能实现工业化嬗变核废料的装置。通过设计1个10 MW的ADS物理方案,研究ADS的嬗变能力。采用MCNPX和ORIGEN的耦合程序,利用基于ENDF6.8处理所得的6个温度(300、600、900、1 200、1 500、1 800 K)下连续能量核数据库,计算得到ADS随燃耗时间变化的有效增殖因数keff、功率峰因子和质子束流强度。同时通过计算给出了该设计方案下ADS燃料多普勒系数、冷却剂空泡系数和有效缓发中子份额,利用这些物理量研究了该ADS方案的安全特性,并通过燃耗计算研究了ADS的嬗变能力。结果表明,在1 000 d燃耗时长内,keff和质子流强随时间的波动较小,燃料燃耗深度较浅,系统可提升功率运行,在假想事故下系统能保持次临界状态。系统嬗变支持比约为8。 相似文献