Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

Research reactor RBT-6

Conclusions Reactor RBT-6 is simple in construction and is easily accessible for conducting experiments. The values of neutron flux in it are high for small thermal power; this together with the large duration of continuous operation ensures the possibility of conducting a wide range of experimental investigationa. Such a reactor may be recommended as a research reactor for irradiation of samples of materials up to moderate flux values (10¹⁹–10²¹ neutrons/cm²) and for conducting experiments for studying the change in the properties of materials during irradiation, which are becoming increasingly more important. Estimates show that the number of used fuel assemblies of the SM-2 reactor are sufficient for the operation of several such reactors.If necessary, the number of experimental channels in the active zone of such a reactor can be increased by increasing the number of fuel assemblies and the thermal power. Beryllium can be used as the lateral reflector. This results in a decrease of the volume of the active zone and the thermal power of the reactor, but increases its cost.Translated from Atomnaya Énergiya, Vol. 43, No. 1, pp. 3–7, July, 1977.     

Thermal–hydraulic feasibility analysis on uprating the HTR–PM

In order to meet energy demand in China, the high temperature gas-cooled reactor–pebble-bed module (HTR–PM) is being developed. It adopts a two-zone core, in which graphite balls are loaded in the central zone and the outer part is fuel ball zone, and couple with a steam cycle. Outer diameter of the reactor core is 4.0 m and height of the core is 9.43 m. The helium inlet and outlet temperature are 250 and 750 °C, respectively. The reactor thermal power is 380 MW. Preliminary studies show that the HTR–PM is feasible technologically and economically. In order to increase the reactor thermal power of the HTR–PM, some efforts have been made. These include increasing the height of reactor core, optimizing the thickness of fuel zone and better selection of the scheme of central graphite zone, etc. Basic design concepts and thermal–hydraulic parameters of the HTR–PM are given. Measures to increase the thermal power are introduced. Thermal–hydraulic analysis results are presented. The results show that, from the viewpoint of thermal–hydraulics, it is possible to increase the reactor power.     

增殖燃烧一体化快堆插花式倒料方案研究

增殖燃烧一体化快堆利用快堆的增殖特性,通过倒料完成从增殖组件向燃烧组件的过渡,从而实现增殖和燃烧过程的一体化。全寿期内燃烧组件提供堆芯的绝大部分功率,而在燃烧组件周围的贫铀组件则将其中的²³⁸U转化为²³⁹Pu,实现增殖功能。通过定期倒料,堆芯在一次装料后可实现长期自持临界,维持几十年的稳定运行。合理的堆芯布置与倒料方案可更好地平衡燃料的燃烧和增殖过程。插花式的堆芯布置与倒料方案是将一部分增殖组件分散布置在堆芯高通量区,保证了增殖组件的快速增殖,同时可保持堆芯在整个反应堆寿期内具有稳定的功率分布。另外,插花式堆芯布置与倒料方案最终的组件卸料燃耗是相对均衡的,所有从燃烧区倒出的组件均具有相近的燃耗,一般在250~300 GW•d/t左右。这使得增殖燃烧一体化快堆可在不进行燃料后处理的条件下,实现铀资源的高效利用。     

Response of fully developed pipe flow to a wall-attached ring-type roughness element

The flow field downstream of a wall-attached roughness element in heated axisymmetric flow has been examined experimentally. Velocity and temperature fluctuations at selected points were examined for sensitivity to changes in position relative to the disturbance. Results indicate that the turbulent Prandtl number in the downstream flow field is in the range 0.4–1.2, that large Reynolds stresses are generated near the tip of the roughness element but diffuse towards the core with downstream distance, and that spectral peaks exist which decay rapidly with downstream distance. The restraining effect of the wall on the lateral turbulence intensity is also well illustrated. The rapidly changing features of the flow with downstream distance from the flow disturbance open up possibilities for monitoring of reactor fuel assemblies as well as in-service location of faults.     

基于钒探测器补偿的轴向偏移控制设计

机械补偿控制是基于控制棒调节反应堆功率并进行轴向功率偏移控制的先进技术,频繁移动的控制棒对传统堆外校准后的轴向功率偏差测量精度有较大的负面影响,采用能给出准确堆内功率分布的钒基自给能探测器的信号可对其进行校正。为补偿钒探测器较长的响应时间,设计了超前/滞后控制器,提出了用补偿后的信号对堆外测量得到的轴向功率偏差信号进行修正的方法。仿真结果表明,该方法能有效应对控制棒移动对轴向偏移控制的影响,可提高控制精度。对该方法的安全相关影响、可实施性及性能分析表明,其具有较高的实用价值。     

Xenon Oscillations in a VVÉR-1000 Core

Xenon oscillations – periodic redistribution of the power over the core volume – can occur in a VVÉR-1000 core because of the large size of this core. The xenon oscillations can be conventionally divided into axial, radial, diametral, and azimuthal. In the present paper, methods are described for initiating the oscillations and the results of experimental investigations of the characteristics of free axial, diametral, and azimuthal xenon oscillations in the reactor core in the No. 1 unit of the Rostov nuclear power plant are presented; the experiments were performed at the start of burnup of the first (head) fuel load, completely consisting, for the first time, of improved fuel assemblies. Among other things, it is established that the periods of all oscillations investigated are the same. The azimuthal xenon oscillations are most rapidly damped.     

高通量工程试验堆考验回路功率分布重组及堆芯热点因子计算研究

利用精细注量率重组和注量率形状因子的乘积方法对高通量工程试验堆的考验回路和堆芯燃料组件内的功率分布进行重组 ,并给出全堆芯的热点因子及其出现的时间 (燃耗步 )、组件位置、轴向位置、径向环位置和环向角度     

Monitoring the distribution of energy release in BN-600 by γ scanning fuel assemblies

A complex of computational and experimental measures for monitoring the distribution of energy release in the core has been perfected over the 25-year history of BN-600 operation in the Beloyarskaya nuclear power plant. Continual monitoring is conducted together with calculations based on three-dimensional multigroup calculations and periodic γ scanning of regular BN-600 fuel assemblies with upgrading of the core of this reactor. At the present time, in the course of switching BN-600 to a new core with four refuelings and maximum fuel burnup increased to 11.1% h.a., the experimental procedure has been upgraded and optimized taking account of the experience gained, three series of such measurements have been completed, and new experimental data on the character of the radial and axial neutron-field distribution have been obtained. Translated from Atomnaya énergiya, Vol. 105, No. 6, pp. 339–344, December, 2008.     

Computational Analysis of Heat Removal from the Core-Melt Catcher in a BN-800 Vessel in the Case of a Serious Unanticipated Accident

A numerical analysis is performed of the cooling efficiency of a bed consisting of fragments of a destroyed core on a BN-800 catcher. A stationary model of the effective thermal conductivity is used to calculate the vertical distribution of the temperature in a heat-releasing layer, including a porous layer located in the coolant, taking account of the aggregate state of the components. The ST0-BED code is tested on numerical results obtained using explicit expressions derived from an analytical solution. The physical accuracy of the method is checked on the results of series-D experiments performed at the Sandia Laboratories in the USA. The numerical estimates show that the cooling of the heat-releasing mass consisting of fuel and the source material of the core assemblies on the BN-800 catcher occurs in the case of a serious accident with heat release density corresponding to 5.5 h after the reactor becomes subcritical. The maximum temperature in the bed at this time will be lower than the boiling temperature of the fuel. The temperature on the catcher is 650–900°C.     

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.     

Design and analysis of a thermal core for a high performance light water reactor

The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25 MPa with a core outlet temperature of 500 °C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 °C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat-up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. The concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility. Coupled neutronic/thermal-hydraulic analyses are defining the initial distribution of enrichment, control rod positions and the use of burnable poisons. Sub-channel analyses predict the coolant mixing inside assemblies, and a porous media approach simulates the flow of moderator water between assembly boxes. Finally, structural analyses of the assembly boxes are needed to minimize deformations during operation. Even though the core design cannot yet considered to be final, this state of the art review shall summarize the progress achieved so far and outline the remaining challenges.     

Determination of Tehran Research Reactor power by N gamma detection

The radioisotope ¹⁶N is produced by the interaction of fast neutrons with ¹⁶O in water reactor coolant. This radioisotope emits at the two major gamma ray energies of 6.13 MeV and 7.1 MeV. Exploiting the linear relation between the number of gamma particles versus the reactor power change, the reactor power is determined by detecting and counting the emitted gammas. In this work, for the detection of gammas to measure the reactor power, two different methods are employed. First, by NaI(Tl) scintillator detector and second, by assembly of ten GM detectors. The obtained results confirm that the number of emitted gammas is proportional to the change in reactor power as shown by different monitoring systems such as UIC, CIC, FC, Cherenkov and thermal power. Both of the applied methods are shown to give reliable results for reactor power above 20 kW. Both systems, having been calibrated, are being used as monitoring systems of power in Tehran Research Reactor. These systems are usable in other research reactors and possibly in power reactors as well.     

Possibilities of Deep Burnup of Americium and Neptunium in the Core of a Fast Sodium Reactor

One-time deep burnup of actinides (up to burnup >90% h.a.) followed by burial in geological formations without reprocessing is of great interest. In this paper, a method is proposed for deep burnup of americium and neptunium in special assemblies, containing actinides in an inert (stone-like) matrix and a strong moderator (zirconium hydride). Placing 130 such assemblies into a BN-800 core with oxide fuel permits 90–95% h.a. burnup in a realistic time – 2–3 runs (2.5–3.5 yr). Such a reactor permits utilizing up to 110 kg of americium and neptunium per year.     

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.     

基于贝叶斯推断的堆芯功率分布重构

基于贝叶斯推断理论,实现了一种有效融合堆内中子探测器实际测量值与中子学理论计算值两类信息的堆芯功率分布重构方法。应用大亚湾核电站1号机组的测量数据对贝叶斯推断方法的功率分布重构精度进行了验证,并将贝叶斯推断方法与卡尔曼滤波方法以及耦合系数法进行了精度对比。验证结果显示,贝叶斯推断方法在整个循环寿期内的均方根误差、最大相对误差、功率峰重构误差分别不大于0.31%、1.64%和0.07%,且重构精度优于卡尔曼滤波方法以及耦合系数法。重构精度以及计算速度表明贝叶斯推断方法有潜力被应用于功率分布在线监测系统。      

Severe accident recriticality analyses (SARA)

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality—both super-prompt power bursts and quasi steady-state power generation—for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45–2000 kg s⁻¹ injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g⁻¹, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s⁻¹. In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady-state power following initial power excursion was in most cases approximately 20% of the nominal reactor power, according to SIMULATE-3K and APROS. However, in some RECRIT cases higher power levels, approaching 50% of the nominal power, were predicted leading to fuel temperatures exceeding the melting point, as a result of insufficient cooling of the fuel. Long-term containment response to recriticality was assessed through MELCOR calculations for the Olkiluoto 1 plant. At a stabilised reactor power of 19% of nominal power, the containment failure due to overpressurisation was predicted to occur 1.3 h after recriticality, if the accident is not mitigated. The SARA studies have clearly shown the sensitivity of recriticality phenomena to thermal-hydraulic modelling, the specifics of accident scenario, such as distribution of boron-carbide, and importance of multi-dimensional kinetics for determination of local power distribution in the core. The results of the project have pointed out the importance of adequate accident management strategies to be used by reactor operators and emergency staff during recovery actions. Recommendations in this area are given in the paper.     

华龙一号“177堆芯”特点分析

华龙一号（HPR1000）压水堆核电厂最显著的技术特征是反应堆采用由177个燃料组件构成的堆芯（简称“177堆芯”）,具有完全的自主知识产权。为深入分析其特点,本文介绍了“177堆芯”的主要技术特征,并在燃料组件及控制棒组件数目方面与157个燃料组件构成的堆芯（简称“157堆芯”）进行了对比分析;对2种典型反应堆堆芯（“177-A堆芯”与“177-B堆芯”）装载方案的异同进行了叙述和评价。结果表明,与“157堆芯”相比,“177堆芯”在安全性和经济性方面更有优势;2种典型堆芯的首循环装载布置各有所长,在可燃毒物选材上,“177-B堆芯”优于“177-A堆芯”。最后,从取消堆芯中央位置控制棒组件、设置堆芯径向金属反射层、实施无中子源启动、分批装载自主化燃料组件以及优化堆芯活性段长度等5个方面给出了HPR1000反应堆堆芯的优化建议。      

Accuracy of nuclear design of fast and thermal neutron coupled core by SRAC

Supercritical-pressure light water cooled fast reactor adopts the blanket fuel assemblies with depleted uranium fuel and zirconium hydride layer in the core for negative coolant void reactivity. Thermal neutrons are generated in the core of fast reactor. It is called “fast and thermal neutron coupled core”. The purpose of the present study is to examine the accuracy of assembly and core calculations including preparation of the macroscopic cross sections with the SRAC code system for “fast and thermal neutron coupled core” in comparison with the Monte Carlo codes, MVP and MVP-BURN. Accuracy of the neutron multiplication factor and coolant void reactivity calculation has been evaluated in four types of cores of different fractions of the blanket fuel assembly with zirconium hydride rods. The conventional analysis is based on the macroscopic cross sections which are prepared with infinite lattice. The conventional SRAC calculation underestimates the neuron multiplication factor for all types of cores. Other findings are that the conventional SRAC calculation overestimates coolant void reactivity for the cores without zirconium hydride rods, and underestimates coolant void reactivity for the core of all blanket fuel assemblies having zirconium hydride rods. To overcome these problems, it has been proposed that the macroscopic cross sections of seed fuel assembly are prepared with the model that a seed fuel assembly is surrounded by blanket fuel assemblies in order to take into account the effects of the surrounding fuel assemblies. Evaluations show that accuracy of the neutron multiplication factor by the SRAC calculation can be improved by the proposed method.     

Neutron-physical and thermohydraulic calculations of VVR-SM with high-and low-enrichment uranium fuel assemblies

The VVR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences of Uzbekistan is being converted from fuel assemblies with high-enrichment uranium (36% ²³⁵U) to fuel assemblies with low-enrichment uranium (19.7% ²³⁵U). During the conversion process consisting of nine cycles, the IRT-3M fuel assemblies with high-enrichment uranium, which are removed at the end of each cycle, will be replaced with IRT-4M fuel assemblies with low-enrichment uranium. This will require increasing the core size up to 20 fuel assemblies and increasing the power of the reactor to 11 MW. The methods used for and the results of neutron-physical calculations (burnup, power distribution, subcriticality), thermohydraulic analysis, and calculations of the kinetic parameters of a stable state are described for a core with high-enrichment uranium, a mixed core, and the first full core with low-enrichment uranium. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 269–273, May, 2008.