The effect of pulsed irradiation on the swelling of 316 stainless steel in fusion reactors

A time-dependent rate theory formulation has been used to study the effects of pulsed irradiation on point defect and void behavior at elevated temperatures. It is found that point defects in pulsed tokamaks, θ-pinchs and inertial confinement fusion reactors (ICFR) display non-steady-state behavior. The pulsed nature of the irradiation has been shown to produce considerable deviations from steady-state void growth behavior at high temperatures (0.3 T_m to 0.5 T_m). In particular, the amount of swelling in the first-wall can be reduced for ICFR pulsing conditions and pulse widths ranging from a nanosecond to a microsecond. The amount of reduction increases with increased pellet yield at a fixed operating temperature, geometry and ICFR plant power output.     

Solvent Extraction Calculation Model for Purex Process in Pulsed Sieve Plate Column

A new calculation code PULCO has been developed to simulate the Purex extraction process in a pulsed column. The model for the PULCO code is based on the mass transfer taking place at the interface between dispersed drops and continuous phase fluid. The following physical quantities of the moving drop are included in the model for each period of drop motion in a pulsed column:

1. Mass transfer coefficient from liquid drop to continuous phase

2. Diameter and velocity of moving drop in the pulsed column

3. Holdup of dispersed and continuous phases in the pulsed column

4. Longitudinal dispersion coefficient in the pulsed column.

To establish the validity of the PULCO code, experiments were performed using a 50 mm diameter pulsed column which had a sieve plate section height of 2 m and plates spaced 50 mm apart. The experiments were performed by CY tests with use of unirradiated U and Pu. The experimental concentration profiles fitted well with the calculated concentration profiles.     

Neutron Nuclear Data Evaluation for Thorium-232

An evaluation was made on the neutron cross sections, resonance parameters and average neutron yield in fission for ²³²Th in the energy range from thermal energy to 20 MeV. The fission and capture cross sections were evaluated on the basis of the experimental data by converting the relative ratio data into cross section values by making use of recent evaluations for reference cross sections. The total cross section was determined from experimental data in the region from 24 keV to 15 MeV and then extrapolated to lower and higher energies by using the optical model whose parameters had been adjusted as so to reproduce the measured data. The elastic and inelastic scattering, (n, 2n) and (n, 3n) reaction cross sections were calculated by means of the statistical model combined with the optical model. A set of resonance parameters were recommended in the energy range below 3.5 keV and average resonance parameters were deduced in the unresolved resonance region. A value of 7.40 b was chosen for the capture cross section at 0.025 eV, and the picket-fence negative-energy levels were introduced so as to reproduce the non-l/v behavior of the capture cross section in the epithermal region.

The results were incorporated in the Japanese Evaluated Nuclear Data Library, Version 2 (JENDL-2). Comparison was made between the present and other evaluations such as ENDF/B-V and possible reasons for the discrepancy were discussed.     

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT

This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria).

The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor.

The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior.

3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.     

A simple novel and fast computational model for the LIVE-L4

The LIVE-L4 test was conducted to investigate the transient and steady state behavior of the molten pool and the crust influenced by different heat generation rates. The main purpose of this work is to develop a simple novel model of the LIVE code to calculate the entire process of the LIVE-L4 test after the melt of KNO₃–NaNO₃ poured into the test vessel. The LIVE code is a transient code and can be used as a fast computational program to calculate the LIVE tests. Natural convection heat transfer in the melt pool, crust behavior, heat conduction in the vessel wall, and radiative heat transfer were all considered in the model of the LIVE code.In the LIVE code, Asfia–Dhir correlations were used to calculate average and local heat transfer coefficients in the melt pool. With the assumption of no considering the composition change of local melt at melt/crust interface, many important parameters, including the melt pool temperature, heat flux distribution along the vessel wall, the thickness of the crust in steady state, and crust growth rate during the test, were calculated and compared with the LIVE-L4 experimental data.The melt pool Nu calculated by the LIVE code is larger than experimental data due to the use of Asfia–Dhir correlation in the LIVE code, which caused the average heat flux through the vessel wall larger than experiment data except the heating phase of 5 kW. It is attributed that the temperature difference between the melt pool temperature and the interface temperature at melt/crust measured in the test is larger than that calculated by the LIVE code due to the constant interface temperature at melt/crust of 284 °C used in the LIVE code. Crust growth rate calculated by the LIVE code was consistent well with the experiment data. Calculation results indicated that the LIVE code could generally predict the main parameters of the melt and crust well during the LIVE-L4 test.     

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O_2−x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of “best estimates” provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.     

行波堆TP-1堆芯热工水力单通道与子通道分析方法研究

以泰拉能源公司提出的钠冷行波堆TP-1为研究对象,通过钠冷行波堆瞬态安全分析程序TAST得到堆芯各组件内冷却剂、包壳和燃料棒的平均温度分布。用子通道分析程序SACOS-Na对TAST计算得到的最热组件进行详细分析计算,得到该组件内冷却剂的温度、压力和流速分布,并得到燃料棒和包壳的温度场。结果表明：单通道与子通道的结合使用能有效提高计算效率,提高反应堆设计的安全性。     

Kinetic Parameter βeff/ℓ Measurement on Low Enriched Uranyl Nitrate Solution with Single Unit Cores (600𝛗;, 280T, 800𝛗;)of STACY

The kinetic parameter β_eff/? of low enriched (10% EU) uranyl nitrate solution was measured by the pulsed neutron source (PNS) method in STACY This measurement was repeated systematically over several uranium concentrations from about 195 gU/l to about 430 gU/l. Used core tanks were two cylindrical tanks whose diameters are 600 mm and 800 mm and one slab tank which has 280 mm thickness and 700 mm width. In this report, experimental data such as solution conditions, critical solution level for each solution condition, subcritical solution levels where measurements were conducted, measured decay time constants of prompt neutron and extrapolated β_eff/? values are described as well as basic principle of the PNS method. The kinetic parameter β_eff/? values were evaluated also by computation with the diffusion code CITATION in SRAC and the nuclear data library JENDL 3.2. Strong linear correlation has been found between kinetic parameter β_eff/? and uranium concentration regardless of differences of reflecting conditions or core tank conditions. Experimentally or computationally evaluated β_eff/? value is about 90 s^-1 at 195 gU/l and about 170 s^-1 at 430 gU/l and both experimental and computational values show good agreement within an error of 3% which is comparable to about 2% uncertainty of measurement.     

The effect of turbulence modeling on hydrogen jet dispersion inside a compartment space using the HYDRAGON code

The mitigation of hydrogen in the containment of nuclear reactor after the Loss of Coolant Accident is essential to preserve the structural reliability of the containment. This paper presents the results of the systematic work done by using the HYDRAGON code to investigate the effect of turbulence models on the concentration distribution of hydrogen and to determine the HYDRAGON code thermal-hydraulic simulation capability during a severe accident at the nuclear power plant. The HYDRAGON code is developed by the Department of Engineering Physics, Tsinghua University, which is an independent research program. The influence of various types of turbulence models, i.e. a standard k ? ? model, a re-normalized group (RNG) k ? ? model, and a realizable k ? ? model were analyzed and the simulation results were compared with the experimental data. When simulation results were compared to experimental data, it was found that, in most compartments, the standard k ? ? model generally yielded reasonable agreement with the experimental results as compared to RNG k ? ? and realizable k ? ? models; however, for probes P7 and P12, better trend was captured by RNG k ? ? and realizable k ? ? models, respectively.     

Thermal-hydraulic behavior of a marine reactor during oscillations

The effect of ship motion, such as heaving and rolling, on the thermal-hydraulic behavior of marine reactors was investigated. The COBRA-IV-I CODE was modified to analyse the thermal-hydraulic performance on the critical heat flux under oscillating acceleration conditions. The critical heat flux in the code was verified experimentally using freon as a comparison. The Critical Heat Flux Ratio (CHFR) at the hottest channel of the PWR subchannel was analysed using the same code. A system code RETRAN-02/MOD2-GRAV was developed by improving RETRAN-02/MOD2 to simulate the thermal hydraulic transient under ship motion. It was verified by comparison using the experimental results of both two-phase natural circulation flow under heaving motion and single-phase natural circulation flow at an inclined attitude. The code was used to analyse reactor plant behavior in the nuclear ship Mutsu. Natural circulation flow during rolling motion was investigated experimentally. The characteristics of loop flow and core flow rates were clarified. The core flow rate correlated well with the Reynolds number of rolling motion.     

中国实验快堆热工参数的自适应BP神经网络预测方法研究

中国实验快堆（CEFR）堆芯的热工参数是否超出限值是评价反应堆安全运行的标准。本文针对燃料包壳最高温度预测问题,通过堆芯子通道分析程序COBRA生成数据样本后,开发基于BP神经网络自适应算法的智能预测程序,对于特定的单盒组件,仅需给出堆芯进口功率和流量,即可实现燃料包壳最高温度的快速准确预测。结果表明,与COBRA相比,在大规模重复性计算的场景下,自开发程序能节约大量计算时间和算力,提高燃料包壳设计和CEFR运行时的操作效率。实验分析得出BP神经网络方法的最大相对误差不超过6%,平均预测相对误差不超过3%,计算效率提升至原程序的300倍,网络模型的预测精度高,且易推广至实验快堆其他参数预测,具有很大的应用前景。     

Self-Limiting Power Excursion Characteristics of Light Water Reactor, (IV)

A strain gage was used for the measurement, of fuel cladding strain generated during pulse operation tests on the Hitachi Training Reactor. In the analysis of the measured strain, two kinds of correction were called for: (1) the fadiation effect on the strain gage and lead wires, and (2) the temperature effect due to the lag of the gage filament temperature behind the true fuel cladding temperature. The experimental axial strain after the two corrections were applied was 781 × 10^?6 cm/cm for the hottest fuel rod in the pulse operation test with an inserted reactivity of 1.20%δk/k. This maximum strain corresponded to 2,169 kg/cm² of thermal stress and 111 cal/cm²·sec of heat flux. These results were obtained under the condition of maximum temperature in the fuel center of 1,200°C and a fuel cladding temperature of 140°C. When the axial strain was calculated with consideration given to the gap or contact conductance between the fuel and its cladding, a reasonable agreement was obtained between the calculation and the experimental results.     

Research on Adaptive BP Neural Network Prediction Method for Thermal Parameters of China Experimental Fast Reactor

Whether the thermal-hydraulic parameters of China Experimental Fast Reactor (CEFR) core exceed the limit is the standard for evaluating the safe operation of the reactor. For the maximum temperature prediction problem of fuel cladding, after generating the data samples by the core sub-channel analysis code COBRA, an intelligent prediction code based on adaptive BP neural network algorithm was developed in the paper. For a specific single-box component, only the core inlet power and mass flow rate were required to achieve fast and accurate prediction of the fuel cladding maximum temperature. Compared with COBRA, in the scenario of large-scale repetitive calculation, self development code can save a lot of calculation time and rescource, and improve the operating efficiency of fuel cladding design and CEFR operation. The experimental analysis shows that the maximum relative error of BP neural network method is less than 6%, the average prediction relative error is less than 3%, and the calculation efficiency is improved to 300 times of the original code. So the prediction accuracy of the network model is high, and self development code is easy to apply to other parameter predictions of the experimental fast reactor.     

A study on the shell wall thinning causes identified through experiment, numerical analysis and ultrasonic test of high-pressure feedwater heater

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which accelerates as the operation progresses. Several nuclear power plants in Korea have undergone this damage around the impingement baffle – installed downstream of the high-pressure turbine extraction steam line – inside numbers 5A and 5B feedwater heaters. At that point, the extracted steam from the high-pressure turbine consists in the form of two-phase fluid at high temperature, high pressure and high velocity. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of number 5 high-pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the downscaled experimental data in an effort to determine root causes of the shell wall thinning of the high-pressure feedwater heaters. The numerical analysis and experimental data were also confirmed by the actual wall thickness measured by ultrasonic tests. From the comparison of the results for the local velocity profiles and the wall thinning measurements, the local velocity component only in the y-direction flowing vertically to the shell wall, and not in the x- and z-directions, was analogous to the wall thinning data.     

Temperature Effects on Reactivity in Light Water Moderated U02 Cores with Soluble Poisons

Experimental and computational studies have been performed on the temperature coefficients of reactivity in light-water moderated and reflected UO₂ cores with soluble poisons such as boron and gadolinium. Experiments were carried out using the Tank-type Critical Assembly (TCA) in Japan Atomic Energy Research Institute (JAERI). Temperature coefficients of the cores with soluble poisons were measured by changing the temperature of the moderator and reflector from the room temperature to about 60°C. The dependence of temperature coefficients on the core configuration and the concentration of soluble poison was studied with the water level worth method. Temperature coefficients were calculated with a diffusion code CITATION included in the SRAC code system and a perturbation code CIPER for comparison with the experimental data. It was found that the temperature coefficients are always negative in the experimental cores (the water to fuel volume ratio (Vm/V_f) of 1.83) containing boron as soluble poison. On the other hand, the temperature coefficients become positive in the cores with gadolinium due to the deviation of the gadolinium absorption cross section from the 1/v law and the neutron spectral shift with the increase in temperature.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (II)

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage Scoping Test. This paper presents the calculated results and discusses, based on the results, on the phenomena that are important for prediction of the thermal response of fuel rods to a severe accident under the partially un-covered core condition. The calculated results are in good agreement with the experimental results. Namely the dry-out time of each elevation and the temperature behavior in both the slow heat-up and rapid temperature excursion processes are well simulated. The analysis indicates that fuel cladding temperature of the upper part of the test bundle would have reached the melting point of ZrO₂ and fuel center line temperature would have reached the melting point of UO₂ during a rapid temperature excursion which was caused by rapid decreasing of the dry-out level and accelerated by zirconium-water reaction in the lower part.     

Numerical Simulation of Extraction Behavior of Americium and Nitric Acid in the TRUEX Process

A numerical simulation code for the transuranium extraction (TRUEX) process was developed to determine the optimum operational conditions for the separation and recovery of transuranium (TRU) elements, such as americium. Equilibrium constants for the americium extraction by octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were determined by the least-squares method using experimental data. The activity coefficients for nitric acid and water were calculated with equations derived from experimental data. The calculated concentration profiles of americium, by using the best available equations, did not agree well with the experimental data in the extraction section. The stage efficiency was added in this code as a parameter to correct the differences between the calculated and the experimental profiles. The calculated concentration profiles using the modified code were in good agreement with the experimental data when 0.9 was used for the stage efficiency.     

Experimental study on convective heat transfer of water flow in a heated tube under natural circulation

This paper reports on an experimental study on transitional heat transfer of water flow in a heated vertical tube under natural circulation conditions. In the experiments the local and average heat transfer coefficients were obtained. The experimental data were compared with the predictions by a forced flow correlation available in the literature. The comparisons show that the Nusselt number value in the fully developed region is about 30% lower than the predictions by the forced flow correlation due to flow laminarization in the layer induced by co-current bulk natural circulation and free convection. By using the Rayleigh number Ra to represent the influence of free convection on heat transfer, the empirical correlations for the calculation of local and average heat transfer behavior in the tube at natural circulation have been developed. The empirical correlations are in good agreement with the experimental data. Based on the experimental results, the effect of the thermal entry-length behavior on heat transfer design in the tube under natural circulation was evaluated.     

基于Gamma全吸收型BaF2探测装置的数据获取系统

为精确测量keV能区的中子俘获反应截面,研制了一套基于Gamma全吸收型BaF₂探测装置的数据获取系统。数据获取系统工作在Linux环境下,应用FlashADC采集探测器的脉冲信号。程序控制、数据存储、在线监控与分析程序基于Root平台。通过测量⁶⁰Co和¹³⁷Cs源,得到了数据获取系统主要性能的测试结果,其满足GTAF装置精确测量中子俘获反应截面的要求。