铅铋自然循环热工水力分析程序开发

铅铋堆内冷却剂的自然循环对于反应堆的正常运行以及事故工况下的堆芯热量导出均至关重要,相关热工水力分析工作对于支持设计及安审均有重要意义。通过对铅铋堆内一回路系统内主要部件,包括堆芯、热交换器、管道等建立热工水力物理模型,开发了适用于铅铋自然循环瞬态过程模拟的热工水力分析程序,并利用铅铋自然循环回路内开展的自然循环启动实验、功率台阶影响实验等的结果进行了程序的初步验证。结果表明,程序计算得到的结果与实验结果符合较好,能够较好模拟铅铋自然循环的瞬态过程。该程序可以为铅铋堆研发过程中自然循环热工水力分析工作提供支持。     

Reactor core thermohydraulic transients with thermohydrodynamic coupling

A study of the reactor core thermohydraulics in an LMFBR has been performed for the strongly coupled thermo-hydrodynamic transients. A numerical method to solve the coupled energy-momentum equations among multichannels in a core is presented and the computer code ORIFS-TRANSIENT has been developed.The results of sample calculations for a flow coastdown transient to natural circulation following a reactor scram in a typical loop-type LMFBR are as follows: (1) the inter-subassembly coolant flow redistribution due to buoyancy forces is significant under the low flow condition, such as natural circulation; (2) the maximum coolant temperature was decreased by about 80°C (corresponding to about 22% in terms of hot channel factor) due to the flow redistribution; (3) due to thermohydrodynamic coupling between upper plenum and other regions, the maximum coolant temperature was decreased by about 9°C; (4) due to inter-subassembly heat redistribution, the maximum coolant temperature was increased by about 7°C.     

Capability of the RELAP5 code to simulate natural circulation behavior in test facilities

Many advanced reactor designs incorporate passive systems mainly to enhance the operational safety and possible elimination of severe accident condition. Some reactors are even designed to remove the nominal fission heat passively by natural circulation without using mechanical pumps e.g. ESBWR, AHWR, CHTR, CAREM, etc. while in most other new reactor concepts, the decay heat is removed passively by natural circulation following the pump trip conditions. The design and safety analysis of these reactors are carried out using the best estimate codes such as RELAP5, TRAC and CATHARE, etc. These best estimate codes have been developed for pumped circulation systems and it is not proven about their adequacy or applicability for natural circulation systems wherein the driving mechanism is completely different. Some of the key phenomena which are difficult to model but are significantly important to assess the natural circulation system performances are – low flow natural circulation mainly because the flow is not fully developed and can be multi-dimensional in nature; flow instabilities; critical heat flux under oscillatory condition; flow stratification particularly in large diameter vessel; thermal stratification in large pools; effect of non-condensable gases on condensation, etc. Though, these best estimate codes use a six equation two-fluid model formulation for the thermal-hydraulic calculation which is considered to be the best representative of two-phase flows, but their accuracies depend on the accuracies of the models for interfacial relationships for mass, energy and momentum transfer which are semi-empirical in nature. The other problem with two-fluid models is the effect of ill-posedness which may cause numerical instability. Besides, the numerical diffusion associated due to truncation of higher order terms can affect the prediction of flow instabilities. All these effects may lead to inability to capture the important physical instability in natural circulation systems and instability characteristics i.e. amplitude and frequency of flow oscillation. In view of this, it is essential to test the capability of these codes to simulate natural circulation behavior under single and two-phase flow conditions before applying them to the future reactor concepts.In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized.     

反应堆内局部自然循环对非能动余热排出的影响

为研究反应堆堆内局部自然循环对非能动余热排出的影响,利用改进的RELAP5/MOD3.2程序对核动力装置及非能动余热排出系统进行数学建模与理论研究,并利用试验数据进行了校核。研究表明：在核动力装置自然循环运行条件下,由于反应堆上封头旁流及反应堆入口漏流通道的存在,在反应堆活性区、上封头、环腔及下腔室之间构成了局部自然循环流动现象;在主回路自然循环能力较弱时,堆内产生的局部自然循环流动占优,反应堆衰变热无法顺利带出。     

The reactor dynamics code DYN3D – models,validation and applications

The article provides an overview of the reactor dynamics code DYN3D. The code comprises various 3D neutron kinetics solvers, a thermal-hydraulics reactor core model and a thermo-mechanical fuel rod model. The implemented models and methods and the capabilities and features of the code are described. Latest developments of models and methods are delineated. An overview on the status of verification and validation is given. Code applications for selected safety analyses are described. Furthermore, multi-physics code couplings to thermal-hydraulic system codes, CFD and sub-channel codes as well as to the fuel performance code TRANSURANUS are outlined. Developments for innovative reactor concepts, in particular Molten Salt Reactor, High Temperature Gas-cooled Reactor and Sodium Fast Reactor are delineated. The management of code maintenance is briefly described. An outlook on further code development is given.     

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

Numerical simulation of natural circulation experiments conducted at the IIST facility

Natural circulation plays an important role in long-term cooling of pressurized water reactors (PWRs) under small break loss-of-coolant accidents. Recently, natural circulation experiments have been conducted at the Institute of Nuclear Energy Research integral system test (IIST) facility, which is used to simulate the Westinghouse three-loop Maanshan PWR. A numerical simulation is presented to investigate the natural circulation phenomena of the IIST facility with the RELAP5/MOD3 code. The calculated results are in good agreement with the experimental data of the single-phase natural circulation both quantitatively and qualitatively. The influences of power level and system pressure on natural circulation can also be predicted by the current model. Based on the two-phase natural circulation data, the calculated flow rate history is similar to that obtained from the experiment.     

Verification study of the FORE-2M nuclear/thermal-hydraulic analysis computer code

The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments.     

超长寿命小型自然循环铅铋快堆堆芯概念设计研究

以提高铅铋快堆的经济性与固有安全性为目标,开展100 MWt超长寿命小型自然循环铅铋快堆SPALLER-100概念设计,在选用PuN-ThN燃料和²⁰⁸Pb-Bi冷却剂的基础上,提出了一种添加固体慢化剂BeO的燃料组件设计方案,开展了堆芯布置研究和控制棒系统设计,分析了堆芯物理特性与稳态自然循环特性。结果表明：在低燃料装载量和小堆芯体积条件下,SPALLER-100堆芯换料周期达32 a,平均卸料燃耗高达210.38 MW·d/kg(HM),整个寿期内的反应性系数均为负值。稳态运行工况下燃料包壳、芯块最大温度均小于安全限值,反应堆具备一回路自然循环能力和一定流量自动分配能力。     

Passive system reliability analysis using the APSRA methodology

In this paper, we present a methodology known as APSRA (Assessment of Passive System ReliAbility) for evaluation of reliability of passive systems. The methodology has been applied to the boiling natural circulation system in the Main Heat Transport System of the Indian AHWR concept. In the APSRA methodology, the passive system reliability is evaluated from the evaluation of the failure probability of the system to carryout the desired function. The methodology first determines the operational characteristics of the system and the failure conditions by assigning a predetermined failure criteria. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the natural circulation performance. Since applicability of the best estimate codes to passive systems are neither proven nor understood enough, APSRA relies more on experimental data for various aspects of natural circulation such as steady-state natural circulation, flow instabilities, CHF under oscillatory condition, etc. APSRA proposes to compare the code predictions with the test data to generate the uncertainties on the failure parameter prediction, which is later considered in the code for accurate prediction of failure surface of the system. Once the failure surface of the system is predicted, the cause of failure is examined through root diagnosis, which occurs mainly due to failure of mechanical components. The failure probability of these components are evaluated through a classical PSA treatment using the generic data. Reliability of the natural circulation system is evaluated from the probability of availability of the components for the success of natural circulation in the system.     

Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10

Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code.Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR.     

一个高效核反应堆热工水力分析方法

提供了一个高效率的核反应堆堆芯热工水力分析方法。以子通道概念为基础，给出了描述堆芯流体流动与传热特性的数学模型和控制方程。文中采用了两相流的滑移流模型，并考虑了过冷沸腾的影响。引入若干补充关系式，用以确定空泡份额、湍流掺混、阻力系数及热力学参数等的大小，与广泛应用的ＣＯＢＲＡ系列程序不同，本文求解的是压力梯度方程而不是关于速度的方程，大大提高了数值求解的稳定性和计算收敛速度。初步的数值结果与实验结果的比较表明。本文提供的方法和程序是令人满意的。     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

自然循环棒束形通道流动不稳定性起始点研究

针对自然循环条件下3×3棒束形通道内流动不稳定性起始点（OFI）进行了实验和RELAP5数值模拟研究。通过对实验数据进行处理,得出了计算自然循环条件下棒束形通道内OFI对应的热流密度的经验关系式,计算的最大相对误差为20.10%。运用驱动力方法分析了OFI的产生原因,计算结果表明：棒束形通道加热段出口处因过冷沸腾产生气泡,使得自然循环冷热段密度差大幅增大,进而使总驱动力增大,最终促使了OFI的产生。RELAP5对于低压自然循环OFI计算适用性好,其对OFI的计算结果较实验结果更不保守。     

非能动安全壳冷却系统空气流道自然循环的比例分析与设计

在考虑建设试验台架经济性的前提下,缩小比例的单项和整体效应试验台架对研究和开发大型先进压水堆核电站及其分析验证程序都具有重要意义。非能动安全壳冷却系统（PCS）壳外空气流道内的自然循环在安全壳非能动冷却性能中发挥着重要的作用。本文从自然循环的数学模型出发,推导出了单项和整体效应试验台架的比例设计方法。在给定壳内热流密度的条件下,通过PCCSAP-3D程序对CAP1400非能动安全壳的2/5比例单项效应试验理想比例台架（ISF）进行模拟。结果表明,本比例分析与设计方法以及在降低高度台架上模拟自然循环是可行的。     

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is hours instead of the 5–6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less—about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented.     

Pressure drop and velocity distribution in rod bundles with spacer grids

The paper contains experimental data and analysis of the pressure drop of turbulent flow through rod bundles. For laminar flow the dependence of the pressure drop on the pitch-to-diameter and wall-to-diameter ratios is discussed on the basis of theoretical analysis. In addition, correlations for the calculation of the pressure loss due to spacer grids are presented and compared with experimental data.Detailed measurements of the velocity distribution in a full bundle of 19 rods are compared with predictions for fully developed turbulent flow. Moreover, detailed measurements of the velocity distributions upstream and downstream of spacer grids typical for LMFBRs are discussed together with the mass flow separation and redistribution between the subchannels. The mass flow distribution found experimentally is compared with the predictions by a subchannel code. The status of experimental knowledge is shown.     

Diana — A fast and high capacity computer code for interchannel coolant mixing in rod arrays

A mathematical model and digital computer program are presented for the subchannel thermal and hydraulic analysis of sodium-cooled fast reactor fuel assemblies. The newly developed FORTRAN-IV computer code ‘DIANA’ is much more useful than many other subchannel mixing analysis codes, especially for large size fuel assemblies which contain more than about 80 subchannels, and for assemblies undergoing swelling and thermal bowing which cause deformed coolant flow ducts, because of high computing speed, reduction of necessary core memory and accurate solution by momentum conservation. Numerical solutions are presented for a deformed rod bundle which contains 179 subchannels.