A linear programming fuel management model for the HTGR

Linear programming was used as the optimization technique to minimize the amount of U-235 used by an example 1160 MWe HTGR during three and six burn periods while simultaneously maximizing the production of U-233. The reactor core was divided into four concentric annular zones; an “out-in” fuel movement technique was used while the fissile loading of the core was held uniform by adjusting the power peaking constraints. The reactor was linearized by holding the neutron flux constant over each of the burn periods. The model was used to consider three cases: Case 1 consisted of three burn periods with no U-233 recycle, Case 2 consisted of six burn periods with no U-233 recycle, and Case 3 consisted of six burn periods with U-233 recycle allowed at the fourth refuelling event. The results indicate that the amount of U-233 produced in the first eight years of operation of the 1160 MWe HTGR will be sufficient to operate the same reactor with no new U-235 fuel for 3 yr hence.     

Real-Time High-Sensitivity Fuel Failure Detection for HTGR

A real-time high-sensitivity fuel failure detection (FFD) method has been developed, where a wire precipitator radiation detector measures noble-gas fission products (FPs) released from a High Temperature Gas-Cooled Reactor (HTGR). By changing the reference counting rate of the precipitator between the normal state and the failed fuel state in real time in response to reactor operation conditions, i.e. reactor power, fuel temperature, coolant-gas flow rate and so on, fuel failure with an extremely low failure fraction (Release-to-Birth ratio <5×10^?6) can be detected. The reference counting rate is obtained by adding an operational tolerance to the background counting rate that is estimated by a diagnostic equation. The diagnostic equation consists of a release equation for estimating the release rate of noble-gas FPs, a gas circulation equation for calculating concentrations of noble-gas FPs in the primary coolant system and a response equation for determining the detection efficiency of the wire precipitator. The feasibility of the method was evaluated by irradiation experiments using gas swept capsules and the Oarai Helium Gas Loop (OGL-1) in the Japan Material Testing Reactor (JMTR). The background counting rate was estimated with an error of about 20% in real time by the diagnostic equation.     

Young's Moduli of Compact Matrices for HTGR Fuel

Young's moduli of matrices of compacted graphite samples were determined from measured values of ultrasonic-wave propagation through the sample at room temperature. The results were analyzed with particular interest directed to the effects of differences in matrix density and graphite orientation on the propagation velocity, from which the modulus was calculated. The characteristics of the matrices were observed by X-ray diffractometry on the crystallite sizes, and the shapes of the graphite powders constituting the main substance of the compact matrices were scrutinized by scanning electron microscope.

The value of Young's modulus was found to increase with binder content in the matrices constituted uniquely of one kind of graphite material, while in the case of composite matrices embodying a mixture of petroleum coke graphite and natural graphite, the modulus decreased with increasing natural graphite content. The velocity of ultrasonic-wave propagation—from which the value of Young's modulus is derived—depends on the density alone in isotropic matrices prepared from milled isotropic coke graphite, whereas in anisotropic matrices embodying needle coke graphite it depends on graphite orientation as well as on density. The contributions of density (p) and of reflective intensity ratio (IR) to the propagation velocity (V) are expressed ΔV=a (Δo) ⁿ and ΔV=b (ΔIR), where the symbol Δ indicates increment, while a, b and n are constants depending upon the characteristics of the compact matrix.     

Above-ground HTGR design for passive decay heat removal

The design of a small high-temperature gas-cooled reactor (HTGR) for passive decay heat removal which could be located deeply underground was proposed previously. In the present work, analogue design analyses of passive decay heat removal for an above-ground long-life small prismatic HTGR was carried out to obtain the conditions for successful decay heat removal by radiation and conduction inside the reactor building, and by radiation and natural cooling by air at the outer surface of the reactor building. Sensitivity analysis of the peak temperatures of both the core and the reactor building after reactor shutdown was performed by changing the physical characteristics of the reactor regions. Enlarging the reactor building was found to be an effective way to reduce the peak reactor building temperature to within its design limit. By using the obtained condition for design parameters, the appropriate sizes of reactor core and reactor building were evaluated for some reactors. Consequently, criticality and burnup analyses for the proposed reactors were performed to confirm the possibility of designing a long-life core for the core size and reactor power which meet the condition of removing decay heat successfully. Using our design, all the reactors with 20 wt% uranium enrichment could be critical for over nine years.     

HTGR颗粒燃料包覆工艺研究

包覆颗粒燃料涂层工艺是高温气冷堆(HTGR)关键技术之一。在研究制备工艺参数对包覆层性能的影响的基础上,确定了制备包覆颗粒燃料的最佳工艺条件,并制备出达到冷态设计要求的 Triso 型包覆颗粒燃料。     

高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

Statistical analysis of certificate data for HTGR microfuel

The characteristics of microfuel with a multilayered coating, which is part of the HTGR spherical fuel elements in the Astra critical stand, are analyzed and generalized on the basis of certificate data using mathematical statistics. The average values of the parameters of microfuel (size of the kernel, coatings and microfuel, material density, uranium dioxide mass, uranium enrichment, impurity content, and others) are found. Their uncertainties are estimated. It is shown that the characteristics of microfuel from different batches do not differ significantly. The results obtained are used to analyze experiments on modeling of the physical characteristics of HTGR, which are under development, on critical assemblies of the Astra stand. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 237–241, October, 2007.     

Corrosion tests on matrix graphites for HTGR pins

Luch Scientific Production Association. Translated from Atomnaya Énergiya, Vol. 73, No. 2, pp. 98-102, August, 1992.     

Simplified Analytical Model for HTGR Core Seismic Response

Large numbers of graphite blocks are involved in the high-temperature gas-cooled reactor (HTGR) core. In the analysis of time history core response, integral time mesh is small so that many integral steps are necessary, because the impact vibration of the core occurs under seismic excitation and the impact between blocks is complete in several 10^?3σ10^?4s. Therefore, time history analysis of the core seismic response requires considerable computational cost. To reduce this cost, a simplified model for core seismic analysis is contrived.

This report describes simplified modeling in the seismic response of HTGR core consisting of graphite blocks. For the model, one degree-of-freedom mass with nonlinear column characteristics is used instead of the stacked column.

Results of the simplified model are compared with those of the experiment and also with a detailed analytical model. The former is in good agreement with both the experimental and the detailed model.     

MLOCA瞬态堆芯水位监测有效性分析

为了验证中国改进型百万千瓦级(CPR1000)核电站在一回路中等破口失水事故(MLOCA)工况下堆芯冷却监视系统(CCMS)测量的有效性,及定量分析控制系统中反应堆冷却剂泵(简称:主泵)状态的2种判定方法导致的水位计算差异,对CCMS测量原理进行了分析。以RELAP5-3D程序对CPR1000机组进行热工水力建模、使用虚拟数字化控制系统(DCS)模拟其控制逻辑,定量计算了在这2种主泵状态判定方法中CCMS输出水位及其误差,并分析了误差产生的原因。结果表明:2种方案都会引入较大的水位误差,结合状态导向法事故处理程序(SOP)分析,可能使操纵员对堆芯水位判断产生一定的误导。     

Development of high temperature structural materials for the HTGR

In accordance with the HTGR program in Japan, a series of R&D for high temperature structural materials in particular with respect to the HTTR design code has been performed in JAERI for more than 20 years. This paper introduces R&D results of the pressure retaining low alloy steel 2 1/4Cr-1Mo and the high temperature structural alloys Hastelloy XR and Ni-Cr-W superalloy for the design code together with some fruits of recent studies.     

高温气冷堆燃料颗粒SiC包覆层性质研究

高温气冷堆（HTGR）燃料颗粒中的SiC包覆层是阻挡裂变产物释放最为关键的一层。本文采用XRD、Raman光谱以及SEM等方法对不同温度下在大内径喷射流化床内通过化学气相沉积法制备的燃料颗粒SiC包覆层进行微观结构分析,研究SiC包覆层在不同制备条件下的微观结构、成分以及密度变化的影响因素。结果发现,在实验设定的MTS（三氯硅烷）浓度范围内,在1 520～1 600 ℃之间均可制备出C、Si等杂质不明显的β-SiC包覆层,密度略有差异。通过微观结构分析发现,SiC包覆层密度的变小主要是由包覆层内微孔引起,且此微孔在包覆层内呈线性分布,同时基本位于某一相同的沉积表面,因此,微孔的生成与颗粒流化状态密切相关。可见,改善流化质量应是下一步工艺改进的主要方向。     

制备HTGR凝胶球的外凝胶工艺工程化问题研究

中核北方核燃料元件有限公司(CNNC)建造的高温气冷堆(HTGR)核燃料元件生产线在采用外凝胶(EGU)工艺制备UO_2核芯时,存在煮胶液沉淀、分散-胶凝过程中胶液流量无法实现精准控制和凝胶球裂口等问题。为解决这些工程化问题,对凝胶球制备工艺和设备进行了优化和改造,并试生产了10批次的UO_2核芯进行验证。结果表明,改进后的生产线可连续稳定的实现工业化生产,UO_2核芯产品合格率超80%。     

HTGR fuel and fuel cycle

This paper describes some of the basic charactritics of the HTGR fuel with emphasis on the 1160 MW(e) plant now being offered commercially by Gulf General Atomic and some of the aspects of the fuel cycle which are unique to the HTGR. The fuel cycle is based on highly enriched (93%) uranium for the initial and the make-up fissile material; thorium for the fertile material, with the bred ²³³U being recycled at the earliest opportunity. The fuel elements consist only of ceramic materials with the thorium/uranium carbides or oxides in the form of coated particles.     

HTGR plant safety design bases

The high temperature gas-cooled reactor (HTGR) has inherent and design safety features that are sifnificant and unique, requiring a number of safety criteria and approaches that differ markedly from other reactor types. This paper briefly reviews the design of HTGR plants that have been built and are being offered in the United States. It then reviews the safety considerations involved in the design of the plants being offered. The unique features, their development, and their effects on safety criteria are described. The design bases of the prestressed concrete reactor vessel (PCRV) are given particular attention. Operating characteristics of the HTGR and plant response to transient conditions are discussed. The design-basis depressurization accident evolution and related HTGR safety requirements are discussed. Characteristics of the HTGR with respect to technical specifications are discussed, with particular emphasis on the PCRV and the core safety limit.     

On-Line Leak Detection Method for OWL-1 Loop by ARX Modeling Using Dewpoint Signals

Model identification technique based on ARX (autoregressive model with exogenous variable) process was applied to dewpoint data recorded at OWL-1 (Oarai Water Loop No. 1) loop cubicle in JMTR (Japan Materials Testing Reactor) and the dynamical interrelationship between the supply and exhaust dewpoints in the ventilation system of the cubicle was empirically determined.

It was shown that the information so derived on the dewpoint dynamics can assist to enhance the sensitivity of leak detection, if it was incorporated into a leak monitoring system for the OWL-1 loop.

A simple digital filter incorporating the dewpoint dynamics was designed in an attempt to develop an efficient leak monitor for the OWL-1 loop. This filter was applied to the dewpoint data recordings during an abnormal leak that had occurred at the OWL-1 loop in the 43 rd cycle of JMTR operation, which demonstrated the effectiveness of the present method for leak detection at its early stage.     

考虑管路中冷却剂动能的反应堆冷却剂泵惰转计算模型

核主泵惰转惯量设计过小,一旦核电站全厂停电会造成核事故,而设计过大会极大地降低机组效率,因此惰转计算模型的准确性对于保证核电站安全和提高机组效率十分重要。本文考虑管路中冷却剂动能对反应堆冷却剂泵惰转过程的影响,通过启-停机过程中功率守恒方程和泵相似定律,推导并建立了考虑管路冷却剂影响的惰转瞬态计算模型,并给出了泵机组惰转惯量和惰转时间的简单计算公式,使计算结果更精确,工程适用范围更广泛,可应用于核工程和非核工程中惰转惯量的精准设计以及惰转时间的精准计算。      

BF3 for Compensation and Additional Emergency Protection in HTGR

Atomic Energy - The purpose of this work is to investigate materials for effective compensation of excess reactivity in high-temperature gas-cooled reactor plants. The possibility of using gaseous...