Asymmetrical Common-Cause Failures Analysis Method Applied in Fusion Reactors

Fusion reactors like the International Thermonuclear Experimental Reactor (ITER) are generally complex systems, which demand reliability analysis. Common-Cause Failures (CCF) are increasingly important in the reliability analysis of these systems because of the widespread redundancy or similar components in them. However, despite the wide research in CCF, there has been little research on the handling of asymmetries of CCF that is inevitable in ITER. A concept of Multi-Common-Cause Failures (MCCF) and its key assumptions are discussed in this paper. On the basis of MCCF and the assumptions, a transition method named Common-Cause Breakdown Structure (CCBS) was designed to manage the asymmetrical CCF. The CCBS method can be easily applied to most fault tree analysis codes because the CCF treated by CCBS can be handled by traditional CCF models. A redundant system example was modeled and calculated in the reliability and probabilistic safety analysis program RiskA developed by FDS Team. The analysis results for water pumps redundant system applied in Tokamak cooling water system show that CCBS method is adequate and effective.     

Neutronic calculations of the U3Si LEU as a dispersed fuel for Low-Power Research Reactors

Usability of the LEU U₃Si dispersed fuel together with the actual UAl₄–Al HEU fuel (mixed core) in Low-Power Research Reactors (LPRRs) (～30 kW) was assessed in this paper. The use of both fuels together (33% HEU and 67% LEU) in LPRRs seems to be achievable from the neutronic point of view. High Initial Excess Reactivity (IER) can be achieved. To maintain the reactor performance in terms of neutron flux value in the internal and external irradiation sites the reactor power needs to be increased to about 32 kW. However the safety margin of the mixed core is smaller in both normal and accidental operation conditions.     

3D MHD Jet in a Non-Uniform Magnetic Field

The purpose of this paper is to present a two-phase 3D magnetohydrodynamics （MHD） flow model that combines the volume of fluid （VOF） method with the technique derived from induced-magnetic-field equations for liquid metal free surface MHD-jet-flow. Analogy between the induced-magnetic-filed equation and the conventional computational fluid dynamics （CFD） equation is made, so that the equation can be conveniently accounted for by CFD. A penalty factor numerical method is introduced in order to force the local divergence-free condition of the magnetic fields and an extension of the void insulating calculation domain is applied to ensure that the induced-magnetic field at its boundaries is null. These simulation results for lithium liquid metal jets under magnetic field configurations of Magnetic Torus （Mtor） and National Spherical Torus Experiment （NSTX） outboard divertor have shown that three dimensional jet can not be annihilated by magnetic braking and its cross-section will deform in such a way that the momentum flux of the jet is conserved. 3D MHD effects from a magnetic field gradient cause return currents to interact with applied magnetic fields and produce unfavorable Lorentz forces. Under 3D applied non-uniform magnetic fields of the divertor, unfavorable Lorentz forces lead to a substantial change in flow pattern and a reduction in flow velocity, with the jet cross-section moving to one side of the jet space. These critical phenomena can not be revealed by 2D models.     

GASFLOW程序以及COM3D程序在反应堆氢气行为分析上的应用

反应堆在事故情况下的氢气风险一直是反应堆安全研究中非常重要的内容。利用氢气风险管理程序GASFLOW计算了反应堆一回路破口事故后安全壳内的氢气分布,对计算结果进行分析。在GASFLOW计算结果的基础上,应用COM3D程序模拟氢气燃烧和爆炸,研究了氢气浓度以及点火位置对火焰扩散的影响。     

Neutron Anisotropic Scattering Effect in Heterogeneous Cell Calculations of Light Water Reactors

A study on the anisotropic scattering effects in heterogeneous square cells of light water reactors has been performed using the characteristics method. It was found that the effects of the anisotropic scattering were relatively large for the MOX fuel cell because of the large neutron current from the moderator to the fuel region and the k _inf value by the P₀ calculation became 0.10–0.16% larger than that by the P₅ calculation. With the transport correction, the k _inf difference from the P₅ calculation became even larger than that from the P₀ calculation and the k _inf value by the transport correction became 0.18–0.25% larger than that by the P₅ calculation for the MOX fuel cell. The transport corrected self-scattering cross sections of the moderator region become smaller than the non-transport corrected ones and the angular flux distribution becomes more anisotropic with the transport correction. Therefore, more neutrons toward the moderator region between the fuel pellets can slow down to the lower energy region with the transport correction. As a result, the k _inf value by the transport correction becomes larger than that by the P₀ calculation, which is opposite effect to that by the P₅ calculation.     

Systems Analysis of Accelerator and Storage Ring Systems for Inertial Fusion

A computerized systems model is being developed to help assess the relative merits of the various accelerator systems that appear able to meet the requirements for inertial fusion with high energy heavy ion beams. Assessment involves calculating the total cost and parameters that are critical to technical feasibility. Current results showing cost trends and important technical parameters for rf linac/storage ring systems using ions of different mass, charge state, and kinetic energy are presented.     

3D thermo-fluid MHD simulation of single straight channel flow in LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) blanket for its DEMO fusion reactor. The mock-up blanket (TBM), using this concept, will be tested in ITER for its tritium breeding and high-grade heat extraction efficiency. In this TBM, pressurized helium is used to remove the heat from first wall, top and bottom plates of TBM. The Pb–Li is used to extract heat from the breeder zones. The flow of Pb–Li with average velocity 0.1 m/s inside the channel can be significantly modified due to MHD effects, which arise because of the presence of strong toroidal magnetic field. A numerical approach is established to capture this flow modification at higher Hartmann numbers (≥20,000). As a validation part of the developed code, MHD phenomenon is studied in 2-D square geometry and numerically obtained velocity profile is compared with available Hunt's analytical results. Thermo-fluid MHD analysis using this code, has been carried out for single rectangular duct of LLCB TBM. The heat transfer has been studied by keeping hot breeders at both sides of the flow channel. The results suggest modification in steady state MHD velocity profile as the liquid flows along the flow length. However, the temperature in various zone remains well within the maximum allowable limit.     

Application of 3D MHD equilibrium calculation to RMP experiments in the J-TEXT tokamak

The application of resonant magnetic perturbation(RMP) coils could break the initial axisymmetry and change the magnetic topology in tokamak systems. To understand the plasma equilibrium response to the RMP fields, three-dimensional(3 D) non-linear magnetohydrodynamics equilibrium calculations have been carried out using the HINT code for an RMP field-penetration experiment on J-TEXT. The HINT code does not assume perfectly nested flux surfaces, and is able to consider directly the change of magnetic topology due to the RMP field penetrations. Correlations between 3 D equilibrium calculations and experimental observations are presented. The magnetic topologies calculated by HINT were compared with the field topologies obtained from a vacuum approximation method. It turns out that the effects of redistribution of plasma pressure and current due to the formation of magnetic islands at various resonant rational surfaces should be considered self-consistently for understanding the change of magnetic structure. Such changes include changes in the shape and size of magnetic islands, and the distribution of stochastic fields around the magnetic islands and at the plasma boundary, which plays an important role for plasma-wall interactions.     

Development of Safety Assessment Method for Plasma Anomaly Events in Fusion Reactors

The objective of the study is to provide a safety assessment method for plasma transients including thermal response of in-vessel components. We developed a plasma physics model for safety analysis which has been implemented in a safety analysis code (SAFALY). The SAFALY code consists of a 0-D plasma dynamics model and a 1-D thermal behavior model of in-vessel components in the thickness direction. The code can treat hydraulic accidents using the results from a hydraulic code and analyze a passive plasma shutdown due to the impurity release from the wall. The overpower events in International Thermonuclear Experimental Reactor (ITER) were investigated, when the fueling rate and confinement improvement changes. The results show no significant damage to the confinement boundary of ITER is expected, as long as the cooling system works normally.     

Hydrogen-Oxygen MHD Generator Applied to Pulsed Power Required in Nuclear Fusion Research

Designs of large test facilities of nuclear fusion research succeeding the current large Tokamaks such as TFTR, JET and JT-60 show that huge pulsed power is required to operate the new test facilities; 700 MW for 10 s to excite poloidal coils. The present paper proposes three steps of application of MHD power generation to fusion to provide such large pulsed power. The first step is to design and construct a small scale MHD generator which excites the Demo poloidal superconducting magnet (SCM) coil being under construction in JAERI. The operating current is 30 kA with the stored energy of 40 MJ. As the working gas of MHD generator, H2-02 combustion product is selected, seeded with 5%K. The second and third steps are to construct an intermediate MHD channel of 100 MWe and a large channel of 800 MWe. Much improved designs are obtained in the present study, compared with the previous designs. For the large 800 MW generator, the maximum magnetic field becomes 3.5 T with the load current of about 100 kA, while the stored energy in the MHD magnet is estimated to be less than 0.5 GJ which is much smaller than 5~8 GJ of planned poloidal coils. The small MHD channel designed for the Demo poloidal coil is 4 m long with the peak field of 1.8 T. The cryogenic magnet can be self-excited within 20 s. The Demo poloidal coil is charged in about 4 s.     

Estimation of Neutronic Performance in a Hybrid Reactor with Regression Analysis

This study presents regression analysis method used for prediction and investigation of neutronic performance in a hybrid reactor using UO₂ fuel and Flibe (Li₂BeF₄) coolant. The ²³⁵U fraction is increased gradually from 0 to 4% stepped by 1% and the ⁶Li fraction within the Flibe coolant is enriched gradually to 30, 60 and 90% from 7.5%. Relations between ²³⁵U fuel fraction and lithium (⁶Li) enrichment are investigated for the estimation of neutronic performance as the tritium breeding ratio (TBR), energy multiplication factor (M), total fission rate (Σ_f), ²³⁸U (n,γ) reaction and fissile fuel breeding (FFB) in the hybrid reactor. Regression analysis by results obtained by using the code (XSDRNPM/SCALE5) for TBR, M, Σ_f, ²³⁸U (n,γ) and FFB are performed. The results of the regression analysis and the values obtained by using the code (XSDRNPM/SCALE5) are compared with respect to the TBR, M, Σ_f, ²³⁸U (n,γ) and FFB of the reactor. The values calculated from the obtained formulations with regression analysis are found to be in good agreement with results obtained by using the code (XSDRNPM/SCALE5). It is observed that the derived equations from regression analysis could provide an accurate computation of the neutronic performances so that these equations could use for the prediction of TBR, M, Σ_f, ²³⁸U (n,γ) and FFB. In addition, correlation matrix is calculated to determine the degree of relationship between variables as TBR, M, Σ_f, ²³⁸U (n,γ) and FFB.     

汉福特场区反应堆安全封存分析

美国汉福特场区的9座生产堆（B、C、D、DR、F、H、KE、KW和N）属于天然铀石墨水冷堆。1993年,美国能源部决定将8座生产堆（B堆除外）进行中间安全贮存（封存）,封存期为75年;B反应堆作为历史博物馆永久保存,并向公众开放。截至2013年,美国已先后成功完成C、F、DR、D、H、N反应堆的安全封存,KE、KW堆的安全封存将在随后几年完成。文章以C反应堆的安全封存为例,介绍了反应堆安全封存的实施等内容。     

MHD Stability Analysis and Flow Controls of Liquid Metal Free Surface Film Flows as Fusion Reactor PFCs

Numerical and experimental investigation results on the magnetohydrodynamics(MHD) film flows along flat and curved bottom surfaces are summarized in this study. A simplified modeling has been developed to study the liquid metal MHD film state, which has been validated by the existing experimental results. Numerical results on how the inlet velocity(V), the chute width(W) and the inlet film thickness(d0) affect the MHD film flow state are obtained. MHD stability analysis results are also provided in this study. The results show that strong magnetic fields make the stable V decrease several times compared to the case with no magnetic field,especially small radial magnetic fields(Bn) will have a significant impact on the MHD film flow state. Based on the above numerical and MHD stability analysis results flow control methods are proposed for flat and curved MHD film flows. For curved film flow we firstly proposed a new multi-layers MHD film flow system with a solid metal mesh to get the stable MHD film flows along the curved bottom surface. Experiments on flat and curved MHD film flows are also carried out and some firstly observed results are achieved.     

Damage Calculations for Devices in the Diagnostic Penetrations of a Fusion Reactor

Degradation of neutron-sensitive devices for fusion diagnostics is calculated as a function of location in ducts through bulk shields. It is shown that single-crystal semiconductor devices, such as n/p silicon diodes, will undergo severe degradation even at the rear of the shield. It will be necessary to replace devices at frequent intervals or withdraw semiconductor parts after a measurement. Device life can be extended by distance, elbows, shortening of exposure. hardening, or a combination thereof.     

Neutronics Calculations in Pellets and Blankets of Laser Fusion Reactor Concept SENRI-I

The neutronic properties of SENRI-I, a reference design of laser fusion reactor proposed by Institute of Engineering, Osaka University, are discussed on the basis of the one-dimensional neutron transport calculations in burning DT plasmas and blankets. The softening of the fusion neutron energy spectrum, the neutron heating and the neutron multiplication are studied and discussed for the compressed DT pellets with various thickness of fuel plasmas and lead or lead-polyethylene tampers.

The neutronic and thermal features in the blanket of the SENRI-I design are also examined. The tritium breeding ratio is high enough (~1.6), depending on the neutron energy spectrum from a pellet. The maximum temperature increase per 1,000 MJ DT fusion reactions is ~3°C in the inner liquid Li layer and ~1.5°C in the stainless steel first wall. A parametric study is also presented on the effect of varying the thickness of the inner Li blanket ΔRⁱ to examine the thickness required for the enough tritium breeding ratio and energy deposition.     

A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can’t withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.     

Time-dependent Neutronic Analysis for High Level Waste Transmutation in a Fusion-driven Transmuter

This study presents time-dependent transmutations of high-level waste (HLW) including minor actinides (MAs) and long-lived fission products (LLFPs) in the fusion-driven transmuter (FDT) that is optimized in terms of the neutronic performance per fusion neutron in our previous study. Its blanket has two different transmutation zones (MA transmutation zone, TZ_MA, and LLFP transmutation zone, TZ_FP), located separately from each other. High burn-up pressured water reactor (PWR)-mixed oxide (MOX) spent fuel is used as HLW. The time-dependent transmutation analyses have been performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m². The effective half-lives of the MA and LLFP nuclides can be shortened significantly in the considered FDT while substantial electricity is produced in situ along the OP.