铅铋冷却快堆含绕丝燃料组件子通道程序开发与验证

由于铅铋冷却剂流动传热现象的复杂性,准确计算铅铋冷却含绕丝燃料组件的冷却剂和包壳温度是液态金属冷却快堆燃料组件热工分析的重点。本文基于集总参数法对守恒方程进行求解,开发了适用于铅铋冷却快堆的子通道分析程序,对液态铅铋在棒束燃料组件中的摩擦阻力模型、湍流交混模型和对流换热模型进行了适用性分析,并对7棒束大涡模拟和19棒束含绕丝传热实验进行了对比验证。结果表明：包壳和冷却剂温度的最大相对误差低于5%。程序能较好完成铅铋冷却含绕丝燃料组件的热工水力计算,可为铅铋冷却快堆设计提供支持。     

Dynamical Error in Measuring VVÉR Coolant Temperature

A computational-experimental method for predicting and estimating the dynamical error in measuring the coolant temperature in VVÉR transient regimes is examined. The method is based on thermal probing of a thermocouple by passing an electric current pulse through the thermocouple wires and detecting and analyzing the transient characteristic using a special algorithm. The special features of the transient characteristics are investigated and an example of estimating the error is presented.     

Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible.     

Evaluation of Thermocouple Fin Effect in Cladding Surface Temperature Measurement during Film Boiling

Thermocouple fin effect on surface temperature measurement of a fuel rod has been studied at elevated wall temperatures under film boiling condition in a reactivity initiated accident (RIA) situation. This paper presents an analytical equation to evaluate temperature drops caused by the thermocouple wires attached to cladding surface. The equation yielded the local temperature drop at measuring point depending on thermocouple diameter, cladding temperature, coolant flow condition and vapor film thickness. The temperature drops by the evaluating equation were shown in cases of free and forced convection conditions. The analytical results were compared with the measured data for various thermocouple sizes, and also with the estimated maximum cladding temperature based on the oxidation layer thickness in the cladding outer surface.

It was concluded that the temperature drops at above 1,000°C in cladding temperature were around 120 and 150°C for 0.2 and 0.3 mm diameter Pt-Pt-Rh thermocouples, respectively, under a stagnant coolant condition. The fin effect increases with the decrease of vapor film thickness such as under forced flow cooling or at near the quenching point.     

Analysis of a Benchmark Experiment on the Hydraulics and Heat Transfer in a Liquid-Metal-Cooled Assembly of Fuel-Element Simulators

A comparative analysis of the experimental data, which were obtained in a benchmark experiment on the thermohydraulics of a model assembly of fuel-element simulators in a flow of sodium-potassium alloy, and calculations performed by specialists, using thermohydraulic codes, from different countries is performed. The model assembly consisted of 25 fuel-element simulators arranged in a square. Russian specialists used the BRS-TVS.R code to perform calculations of the benchmark experiment, Japanese specialists used SPIRAL and AQUA, Spanish specialists used FLUENT, Dutch specialists used STAR-CD, and South Korean specialists used MATRA and CFX. The following experimental and computational parameters were compared: the coolant temperature in the channels under nonuniform geometric and thermal conditions in the assembly, the surface temperature of the measuring fuel-element simulator on the heated section, and the coolant velocity in the assembly cells around the measuring simulator. Special attention was given to investigating the influence of the spacing lattice on the coolant velocity. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 336–348, November, 2005.     

Soft-measuring method of the core exit temperature of coolant for the 200 MW nuclear heating reactor

This paper reports a soft-measuring method of the core exit temperature of coolant for Chinese 200 MW nuclear heating reactor (NHR-200). The primary and vice sheath thermocouple are immersed in a space orthogonal slot that is located at one side of the support grid plate for fuel assemblies. The core exit temperature of the coolant is evaluated by using these two thermocouple's measurement temperatures. The experimental study gives the formula for evaluating the core exit temperature of the coolant. The space orthogonal slot is smooth to decrease the hydraulic resistance of the coolant, thus a part of coolant was steady flowing through the space orthogonal slot on the support grid plate. So the coolant temperature difference between the center region of fuel assembly and the measurement end of the primary sheath thermocouple is small. The flow of the coolant in the slot increases also the sensitive length of temperature of the sheath thermocouple. All of these ensure the measurement reliability and accuracy. The maximum measurement error of the core exit temperature of the coolant for the NHR-200 is 1.7 K.     

Particulars of the simulation of experiments on supercritical pressure water using other media

Post-reactor investigations have been performed on BN-600 fuel-element cladding, made of 0Kh16N15M3BR steel, after irradiation to maximum burnup of 10% h.a. and higher. It is shown that the highest degradation of the operating properties of the fuel-element cladding is observed in the zone of maximum increase of its diameter and is expressed as total embrittlement of the cladding material and appearance of cracks of substantial depth on the inner surface. The processes resulting in the degradation of the properties of fuel-element cladding are directly related either with swelling or with radiation-induced segregation, occurring in the same temperature range and under the action of the same driving forces as swelling. The most important stresses, from the standpoint of the serviceability of fuel elements, turn out to be those arising in cladding as a result of the gradient of the swelling along the thickness of the cladding. The level of these stresses is also determined by the form of the temperature dependence of the swelling of the steel used for the fuel-element cladding. Translated from Atomnaya énergiya, Vol. 106, No. 4, pp. 188–195, April, 2009.     

Systematic approach to predicting corrosion of zirconium alloys in the water coolant of nuclear reactors

A method of semiempirical prediction of corrosion of cladding zirconium alloys as a function of the operating conditions and composition is presented. The laws of thermodynamics and chemical kinetics of the oxidation reactions of a multicomponent zirconium alloy form the physicochemical basis of the computational method. The method is based on a model developed at the All-Russia Research and Design Institute of Integrated Power Technology for the corrosion of commercial and experimental zirconium alloys in water media under autoclave and reactor conditions taking account of the composition of the alloy and the water chemistry. The model is verified on the basis of independent tests performed on a series of zirconium alloys under autoclave and reactor conditions. The method developed makes it possible to predict the corrosion of fuel-element cladding made from zirconium alloys with fuel burnup to 80 MW·days/kg under the conditions of one- and two-phase VVER and RBMK coolant.     

An analytical method of evaluating the effect of rod displacement on circumferential temperature and heat flux distributions in turbulent flow through unbaffled rod bundles

An analytical method of evaluating the circumferential variations of temperature and heat flux fields inside and around a displaced fuel rod in triangular rod bundles in turbulent flow is presented with illustrative examples. The analysis consists mainly of the derivation of the simultaneous solutions of a set of heat conduction equations for fuel, cladding and coolant under the assumption of fully developed flow and heat transfer conditions. The local coolant velocity distribution, which is necessary for deriving the temperature field in coolant, is determined by solving the Navier-Stokes equation and the turbulent mixing of coolant is taken into consideration. The results show how the circumferential variations in the temperature and heat flux fields on the outer surface of the cladding increase the lower the ratio and the larger the fuel rod displacement due to thermal conduction and peripheral coolant flow velocity distribution.     

Methods for protecting fuel-element cladding from corrosion in lead coolant and the heat-transfer sublayer

The results of investigations of the corrosion of commercial and experimental steels in lead and the possibilities of corrosion protection are presented. The effect of lead coolant and the lead heat-transfer sublayer on fuel-element cladding are examined. Methods based on thermodynamic calculations and experimental data are proposed for protecting fuel element cladding in a lead-cooled reactor from the corrosive effect of the coolant by creating a new corrosion resistant chromium steel and from the corrosive effect of the heat-transfer sublayer by alloying with the components of steel. The results of this work have been implemented in the experimental fuel elements for the BREST-OD-300 reactor which were irradiated in a BOR-60 reactor. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 88–94, February, 2008.     

Method for evaluating the service life of VVER-1000 fuel-element cladding in different loading regimes

A method for evaluating the service life of VVER-1000 fuel-element cladding in constant and variable fuel load regimes is proposed. It is shown that a comparative assessment of the service life of cladding with different core regimes and parameters is possible on the basis of the energy approach to creep and destruction of cladding.     

Zircaloy-U02 and -Water Reactions and Cladding Temperature Estimation for Rapidly-Heated Fuel Rods under an RIA Condition

Zircaloy cladding chemical reactions with coolant water and UO₂ fuel at elevated temperatures under a reactivity initiated accident (RIA) condition were studied from a metallurgical point of view on the basis of the nuclear safety research reactor (NSRR) experiments. The cladding-fuel chemical reaction was extensively analyzed and found to be explainable from equilibrium phase diagrams. The systematic estimation methods of maximum cladding temperature were proposed and examined from metallographies. Maximum cladding temperature can be estimated from measured oxidation thicknesses in the temperature range of 1,000~1,600°C, from melting microstructures in the range of 1,600~1,950°C and also from the volume fraction of the precipitates, (U, Zr)0_2-x, in once-molten oxygen-stabilized α-zircaloy in 1,950~2,400°C. The estimation by the method proposed in the paper is more valid than thermocouple indications at high temperatures, since thermocouples perturb the temperatures they are measuring or fail at the extremely high temperatures. The results are thought to be applicable also to understand general fuel rod behavior under hypothetical accident conditions which cause severe fuel damage.     

Subchannel analysis of a fast reactor cooled by supercritical light water

The concept of a high temperature fast reactor cooled by supercritical water (SCFR-H) was developed for achieving high thermal efficiency and a compact reactor system. The core characteristics were obtained from single channel thermal-hydraulic analysis. Thus, it is necessary to carry out subchannel analysis to estimate the effect of local power peaking and cross flows. For this purpose, a subchannel analysis code is developed. It is verified by comparing the results with experimental data of High Conversion Pressurized Water Reactor (HCPWR). Sensitivities of the outlet coolant and cladding temperature to the subchannel flow area and local power peaking are high. One of the reasons is that the ratio of the coolant flow rate of SCFR-H to the power is smaller than that of LWR. Another reason is that, temperature of supercritical water is more sensitive to the enthalpy change above 450°C. The outlet coolant temperature distribution can be flattened by reducing the area of the peripheral subchannels and by enhancing the mixing between the subchannels.     

Thermo-hydraulic analysis for SCWR during power-raising phase of startup

The study of thermal characteristics during startup is one of the most important aspects for safety analysis of supercritical water-cooled reactor(SCWR).According to the given sliding pressure mode of SCWR,thermal analysis on temperature-raising phase and power-raising phase of startup are carried out.Considering the radial heterogeneity of power distribution,thermal characteristics for different assemblies during startup are also put forward.The results show that,during temperature-raising phase with core power increased only,the temperature of moderator,coolant and fuel cladding in inner assemblies are increased with little amplitude.During power-raising phase with core power and feed-water flow rate increased,the coolant temperature keeps unchanged,but the moderator temperature is decreased.With a greater variation of power,fuel cladding temperature shows a greater increase.Furthermore,considering the uneven distribution of radial power,thermo-hydraulic characteristics with uneven cladding temperature distribution shows a certain horizontal heterogeneity for different fuel assemblies,which becomes serious as flow rate and power increase.By adjusting flow rate distribution in different fuel assemblies or changing power setting during startup,the cladding temperature difference could be effectively reduced,which provides a certain reference for startup optimization of SCWR.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly

In sodium cooled liquid metal reactors design limits are imposed on the maximum temperatures of the cladding and fuel pins. Thus an accurate prediction of the core coolant/fuel temperature distribution is essential to LMR core thermal hydraulic design. The detailed subchannel thermal hydraulic analysis code MATRA-LMR is being developed for LMFBR core design and analysis based on COBRA-IV-I and MATRA. The major modifications and improvements implemented in MATRA-LMR are as follows: sodium property calculation subprogram, sodium coolant heat transfer correlations, and most recent pressure drop correlations. To assess the development status of this code, benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR were compared to the measurements and to the SABRE4 and SLTHEN code calculation results, respectively. Finally, the major technical results of the conceptual design for the KALIMER U-10%Zr binary alloy fueled core have been compared with the calculations of the MATRA-LMR, SABRE4 and SLTHEN codes.     

200MW核供热堆堆芯燃料元件盒冷却剂出口温度测量方法实验研究

提出了将200MW核供热堆芯燃料元件盒支承格子板的一个侧面设计成空交沟槽结构,并在其中放置铠装热电偶测量堆芯燃料元件盒冷却剂出口温度的设计方案。经实验表明,支承格子板侧面空间正交沟槽结构,不仅能有效地导流冷却剂,而且增长了铠装热电偶的感温长度,减小了测导热损失误差,从而提高了测量精度。因此,其测量方法能满足工程上对温度测量精度的要求。     

Single-pass core design of a low-temperature Super LWR

A supercritical-pressure light water cooled and moderated reactor (Super LWR) with a single-pass flow scheme is developed for simplifying upper core structures. Both coolant in the fuel channels and the water rods flow upward and are mixed in the upper plenum. It eliminates the moderator guide/distribution tubes in the upper core that were used in the previous Super LWR design adopting two-pass coolant flow scheme. This core design adopts a four-batch fuel management scheme and an out–in fuel loading pattern. One hundred and twenty-one fuel assemblies with an active height of 3.7 m are included. The flow rate fraction for water rods is 3.5%, and the thermal insulator is used to keep the moderator temperature below pseudocritical temperature. The equilibrium core is analyzed by using neutronic and thermal-hydraulic coupled calculation. The results show that the maximum cladding surface temperature (MCST) is limited to 485 °C with the average outlet temperature of 400 °C. The inherent safety is fulfilled by the positive water density reactivity coefficient and sufficient shutdown margin. On the other hand, the investigation of average outlet coolant temperature varying with MCST is carried out to explore the maximum outlet temperature by employing current MCST criterion and single-pass core design. The average outlet temperature increases with the MCST, and it achieves 465 °C with the thermal efficiency of 43.1% at the MCST criterion of 650 °C. The structure inside the reactor pressure vessel is simplified as a pressurized water reactor.     

RBMK Safety Investigation for Accidents Initiated by Partial Breaks in the Circulation Loop

The present status of RBMK safety investigations for accidents initiated by partial breaks in the head part of the circulation loop is analyzed. The analysis shows that the RELAP code, which is the main tool for simulating such accidents, is inapplicable. An estimate of the maximum temperature of the fuel-element cladding in an accident situation with critical partial rupture is estimated on the basis of an analytical analysis of the thermal conductivity, RELAP5/MOD3.2 code calculations of a hypothetical RBMK-1000 accident, and taking account of the uncertainty of the experimental data accumulated in the investigation of similar PWR accidents. It is found that this temperature can exceed the applicability criteria.     

Core design of a high-temperature fast reactor cooled by supercritical light water

A high-temperature large fast reactor cooled by supercritical water (SCFR-H) is designed for assessing its technical feasibility and potential economical improvement. The coolant system is once-through, direct cycle where whole core coolant flows to the turbine. The goal is to achieve the high coolant outlet temperature over 500°C. We study the reactors with blankets cooled by ascending and descending flow. SCFR-H adopts a radial heterogeneous core with zirconium-hydride layers between the driver core and the blankets for making coolant void reactivity negative. The coolant outlet temperature of the core with blankets cooled by ascending flow is low, 467°C. The reasons are as follows: (1) the power swing due to the accumulation of fissile material in the inner blankets with burn-up, and (2) local power peak in the assemblies due to the zirconium-hydride layers. The difference in the outlet coolant temperature is more enhanced than the low temperature core where outlet temperature is approximately 400°C. The reason is that the coolant temperature is more sensitive to the enthalpy change than near the pseudo critical temperature, 385°C at 25 MPa. Thus, we design the core with blankets cooled by descending flow to obtain high coolant outlet temperature. The coolant outlet temperature becomes 537°C, which is 70°C higher than that of the core with ascending blanket flow. The thermal efficiency is improved from 43.2 to 44.6%. The coolant mass flow rate per electric power decreases by 14%. This will reduce the size of the balance of plant (BOP) system. The power of the reactor is high (1565 MWe) and the void reactivity is negative.