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.     

Change in shape and size of fuel assembly duct in BN-350 reactor

The paper presents analytical data obtained by directly measuring samples taken at different levels of a hexagonal fuel assembly duct used in the BN-350 fast reactor. The difference between the length of the deflection arc and the width of an intact duct face displays the accumulated irradiation-creep deformation during the operation period. The deflection of the duct faces induced by the irradiation creep was analyzed, and irradiation-creep rate versus distance from the core center was analyzed for the assembly duct in the BN-350 fast reactor.     

零功率反应堆内中子通量相对分布实时测量

在堆物理实验中,经常需要进行堆内中子通量相对分布的测量,以便获得有关的参数,如全堆平均热中子通量及功率不利因子、控制棒对中子通量分布的影响等等。为了要获得这些数据,有时不得不进行几千个测点的测量,才能求得结果。以往一般都采用经典的活化法。这种方法的最大缺点是测量工作量大,花费的人力多,不能很快地得到所需要的结果。为此,我们利用一种微型的中子探头,配以适当的电子仪器和机械设备,在轻水零功率反应堆内进     

MOX燃料在轻水堆核电站中的应用

目前MOX燃料已成为一种可用于轻水堆核电站成熟的核燃料。简要介绍了国外该领域的发展状况以及MOX燃料对反应堆性能的主要影响和应对措施。探讨了MOX燃料在国内压水堆核电站中的应用问题。     

国产反应堆压力容器的辐照脆化行为及预测

反应堆压力容器（RPV）是保障核电站运行安全性、经济性的核心构件。对RPV的完整性评估而言辐照脆化是必须面对的问题。我国已开发了第三代设计寿命为60 a的核电站。当达到寿期末时,辐照脆化的行为是未知的,这给国产RPV的辐照脆化预测带来了困难。为研究高注量下的辐照脆化行为,对A508-3钢的材料力学性能试样进行辐照考验,辐照温度为(288±8) ℃,中子注量水平达到反应堆压力容器60 a寿期末的辐照水平1×10²⁰ cm^-2;开展拉伸、冲击和断裂韧性试验,分析辐照脆化行为,在EONY模型基础上,提出针对国产RPV钢的改进的辐照脆化模型。模型的有效性被试验数据证实,其可准确预测国内A508-3材料的辐照脆化趋势。     

国产A508-3钢辐照性能

反应堆压力容器（RPV）作为反应堆寿期内不可更换的核心设备,是防止堆芯放射性泄漏的最主要屏障。本文针对国产压力容器材料A508-3钢,开展了一定剂量水平（约10×10¹⁹ cm^-2,E≥1 MeV）的研究堆加速辐照试验,并进行了辐照后力学性能测试分析,包括拉伸性能和冲击性能测试。结果显示,辐照后在-100、20、288 ℃下,A508-3钢的屈服强度分别增加了83、108、52 MPa,抗拉强度分别增加了58、61、49 MPa,韧脆转变温度T_41J增加了68 ℃,上平台能量降低了61 J。A508-3钢辐照前后性能测试结果表明,在中子辐照至60 a寿期后,A508-3钢仍能满足反应堆使用要求。     

On the diffusion coefficient calculation in two-step light water reactor core analysis

This paper presents consistent and rigorous accuracy assessments of various methods for calculating the diffusion coefficients in a two-step reactor core analysis of light water reactors (LWRs). The diffusion coefficients are significantly affected by the transport correction and critical spectrum calculations. There are various methods for the transport corrections (inflow/outflow/hybrid corrections) and critical spectrum calculations (B₁/P₁/CASMO-4E methods) so that it is necessary to decide the best combination to achieve a high accuracy in the transport/diffusion two-step analysis. Numerical tests are performed step-by-step to search for the best combination of the methods by comparing each other the transport one-step results, transport/diffusion two-step results, and Monte Carlo results. Numerical test results with a large and a small LWR core show that the combination of inflow transport correction and CASMO-4E critical spectrum calculation is most accurate than the other combinations in terms of eigenvalues and assembly power distributions.     

A Multicomponent Model of Radiation-induced Segregation for Commercial Stainless Steels

A radiation-induced segregation model of the Fe-Cr-Ni-Si-Mo system was developed to predict the change in grain boundary composition in commercial stainless steels under neutron irradiation. Parametric survey showed that there were strong influences on segregation behavior for the binding of Si to an interstitial and the jump frequency of Ni and Si via interstitial mechanisms. Careful setting of these parameters using effective point defect migration energies resulted in a good agreement between calculated results and measured data for SUS316 stainless steels irradiated to 74 dpa in light water reactor. Model calculation qualitatively reproduced the effects of dose rate, temperature, and bulk composition reported in the literature.     

Heterogeneity of ion irradiation-induced hardening in A533B reactor pressure vessel model alloys

Reactor pressure vessels comprise bainitic steel structures, and are heterogeneous on the mesoscale. Nanoindentation techniques were used to evaluate the hardness of these structures on the micrometer scale, and to evaluate the heterogeneity in a specimen using the distribution of the hardness. Three A533B model alloys were irradiated by 2.8 MeV Fe²⁺ ions at 563 K, and the effects of ion fluence, ion flux, and chemical composition on the change in the hardness distribution were examined. Heterogeneity of the hardening is observed in high-copper specimens irradiated up to (2–10) × 10¹⁴ ions/cm², where the average hardness increases the most. In these specimens, the hardness distribution broadens, and demonstrates that the hardening in certain positions (possibly where the initial hardness is high) is greater than in other positions. Variation in initial chemical composition (especially copper and carbon) or sink strength may cause a difference in the curing behavior.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking

Isolation of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) was attempted by means of low-dose high-temperature neutron irradiation in a material test reactor to get better understanding on IASCC. Microstructure, grain boundary segregation, hardness and SCC susceptibility were examined on stainless steels irradiated to 0.8 dpa at around 673 K. The irradiation caused well-developed grain boundary segregation without notable hardening or microstructural changes. No IASCC was found in 593 K hydrogenated water whereas the steels were highly susceptible to IASCC in 561 K oxygenated water. The results suggest that grain boundary segregation, probably Cr depletion, is sufficient to cause IASCC in oxygenated water and that other radiation-induced changes such as microstructure and hardening are required for IASCC in hydrogenated water.     

Application of homogenization techniques for inflow transport approximation on light water reactor analysis

The transport cross-section based on inflow transport approximation can significantly improve the accuracy of light water reactor （LWR） analysis,especially for the treatment of the anisotropic scattering effect.The previous inflow transport approximation is based on the moderator cross-section and normalized fission source,which is approximated using transport theory.Although the accuracy of reactivity is increased,the P₀ flux moment has a large error in the Monte Carlo code.In this s...     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

Dissolution behavior of core structure materials by molten corium in boiling water reactor plants during severe accidents

For the decommissioning of the Fukushima Daiichi Nuclear Power Plant, it is necessary to consider the access route to the fuel debris for its removal, which can be determined by knowing the corruption situation of the core support structure. To predict the damage condition of reactor vessel, dissolution behavior of the core structure material should be understood. In this study, the dissolution behavior of core structure materials (stainless steel) by molten metallic corium (stainless steel + B₄C) originated from control rod and its cladding was investigated. As a result of immersion experiment, it was found that there were two types of dissolution mode in this system: (1) chemical dissolution by eutectic reaction between Fe and B and (2) physical dissolution caused by the grains falling off from solid steel due to infiltration of molten metal. Moreover, on the basis of kinetic analysis, it was considered that the chemical dissolution in this system was slow. Therefore, the dissolution is considered to mainly occur through the mechanism that physical dissolution precedes chemical dissolution.     

Analysis of fuel temperature effects on reactivity of light water reactor fuel assemblies by using MVP-2 adopting an exact resonance elastic scattering model

Using the continuous-energy Monte Carlo code MVP-2 adopting a resonance elastic scattering model considering the thermal motion of a target nucleus (the exact model) for major heavy nuclides, analysis of fuel temperature effects on reactivity of mockup UO₂ and MOX fuel assemblies for light water reactors was performed, and the results were compared with those of the conventional asymptotic model. A base condition was a hot operating condition with an in-channel void fraction of 40% and fuel temperature of 520 ℃ for the BWR fuel assemblies and a hot zero-power condition with fuel temperature of 284 ℃ for the PWR fuel assemblies. The fuel temperature of a high-temperature condition was 1500 ℃ for both types of assemblies. The calculated results showed that the exact model made the neutron multiplication factors at the high-temperature condition lower by ?220 to ?440 pcm (10^?5 Δk) and the Doppler reactivity between the base- and high-temperature conditions more negative by 7% to 10% compared with those obtained by the asymptotic model. The energy-dependent reaction rates of capture and ν-fission were also analyzed to study the detail mechanism in the effect of the exact model on the assembly reactivity.     

Effects of Dissolved Hydrogen and Strain Rate on IASCC Behavior in Highly Irradiated Stainless Steels

IASCC behavior in cold-worked SUS316 stainless steels irradiated to 35 dpa was examined using slow strain rate tensile testing at a strain rate of 6.7 × 10^?8/s in 320°C simulated PWR primary water while varying the dissolved hydrogen (DH) concentration from 0 to 2.8 ppm. The results were compared with those previously obtained at a higher strain rate using specimens of different sizes and with those of the previous interrupted experiment. The initiation and propagation of IASCC enhanced with increasing DH concentration and lower strain rate. The IASCC initiation stress decreased to almost half of the yield strength at high DH. Accompanying slow tensile tests in an argon gas environment showed that a lower strain rate did not change in the initiation stress that exceeded the yield strength, but enhanced the propagation of intergranular cracking.     

Damage region formation in ferritic steels induced by light water reactor neutron spectrum

A new theoretical model for damage region formation is proposed. The model is based on numerical solution of the Boltzmann transport equation for knocked-on atoms. A key point of this model is the selfconsistent determination of subcascade overlapping energy E_over (the threshold energy for distinguished damage region formation). Damage region density and size distributions in ferritic steels (Fe–0.2 wt% Cu and Fe–0.2 wt% Cu–0.3 wt% Si) under neutron irradiation in light water reactor spectrum were calculated.     

Research on structural integrity of pressure boundary in light water reactor designs

This paper is a review of the recent researches performed and planned in Japan relevant to the structural integrity of the pressure boundary in light water reactor designs. Various aspects of relevant work on materials, pressure vessel and piping models are described.     

医院中子照射器的反应堆水净化系统抗震分析

医院中子照射器堆水净化系统实际设计等级为3级,有抗震要求,抗震设计标准为TECDOC-1347规范,可以使用等效静力法对堆水净化系统进行抗震计算.首先根据TECDOC-1347规范,说明如何取抗震计算参数.然后根据所取的参数,对整体结构进行了抗震计算.计算中直接考虑了螺栓联接,避免了手动计算螺栓强度.根据TECDOC-1347的要求,要对结构进行屈曲分析.在软件中建立的三维计算模型,忽略螺栓的影响,计算了整体结构的临界屈曲载荷.通过分析证明,结构能够满足抗震要求,在地震载荷和内压的作用下,结构不会发生屈曲.     

Characterization of solid particles sampled from condensates in boiling water reactor

The growth, activation and deposition of corrosion products are the primary sources of radiation buildup on the surface of out-of-core piping in nuclear power plants. The buildup of radiation can have negative effects on the performance of the facility and cause harm to staff during maintenance outages for refueling. This paper reports on the crystalline and amorphous structures of corrosion products sampled in the boiling water reactors in nuclear power plants of Kuo-Sheng and identified using an acid dissolving technique. X-ray diffraction, scanning electron microprobe and inductively coupled plasmaatomic emission spectroscopy were used to analyze the samples. The results indicate that the quantity of amorphous iron oxide at inlet of the condensate demineralizer in Unit 2 is higher than that in Unit 1. The proportion of crystalline to amorphous corrosion products can affect the efficiency of removal. Thus, these results can be used to explain the difference in removal efficiency of condensate demineralizers in different units. Moreover, the iron oxide structures with various properties were observed in different operational periods. It is probable that the higher proportion of amorphous structures with a smaller particle size would reduce efficiency in the removal of condensate demineralization in Unit 2.