Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes.Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results.Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests.The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW.The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.     

Indication Lowering of Average Magnitude Type Campbelling System in Low Pulse Rate Region

In the nuclear power calibration process of JOYO, the first experimental fast breeder reactor in Japan, the indication of the Intermediate Range Monitoring System (IRMS), employing Average Magnitude type Campbelling System (AMCS), was calibrated with the reactor power at 45.82 kW. The reactor power was then decreased and the nonlinear indication lowering of the IRMS was observed.

In this paper, we present a derivation of an equation representing the AMCS characteristic and show that the indication lowering occurs when the product of mean arrival rate of pulses and their width is small. The computed values based on the derived equation agreed very well with the observed ones in the JOYO IRMS below 45.82 kW, and it was confirmed that the evaluation method was applicable. Furthermore, it was evident from the evaluation that the indication lowering of the JOYO IRMS above 45.82 kW was negligibly small, and did not affect the reactor power ascension testing. Also, it was proved in a following thermal calibration test above 50 kW that the indication lowering was allowably small indeed in this power region.     

Thermal-hydraulic behavior of the NUR nuclear research reactor during a fast loss of flow transient

Nuclear safety analysis remains of crucial importance for both the design and the operation of nuclear reactors. Safety analysis usually entails the simulation of several selected postulated accidents, which can be divided into two main categories, namely reactivity insertion accident (RIA) and loss of flow accident (LOFA). In this paper, thermal-hydraulic simulations of fast LOFA accident were carried out on the new core configuration of the material test research reactor NUR. For this purpose, the nuclear reactor analysis PARET code was used to determine the reactor performance by calculating the reactor power, the reactivity and the temperatures of different components (fuel, clad and coolant) as a function of time. It was observed that during the transient the maximum clad temperature remained well below the critical temperature limit of 110 °C, and the maximum coolant temperature did not exceed the onset of nucleate boiling point of 120 °C. It is concluded that the reactor can be operated at full power level with sufficient safety margins with regard to such kind of transients.     

Gallium-cooled liquid metallic-fueled fast reactor

We have examined the effects on core characteristics of using two different types of Pu-based metallic alloy fuels in the gallium-cooled fast reactor core. In the proposed concept, the liquid metal fast nuclear reactor uses metallic fuel in the liquid phase and gallium coolant at high temperature (inlet 1700K, outlet 1900K). The liquid fuel is continuously supplied to the reactor during operation at full reactor power. The reactor power is controlled by rotational control drums with absorber material. The aim was to evaluate reactor core neutronics and safety characteristics demonstrating a feasibility of the reactor system. Although gallium has large absorption cross section in the high neutron energy region, we can design the core with rather good neutronics performances. The large negative reactivity feedback induced by the thermal expansion of liquid metallic fuel ensures the core's inherent safety against the unprotected loss-of-flow transient.     

Transient Safety Analysis of Small Leaded-Bismuth Cooled Fast Reactor

Based on the multi-channel thermal model and the power model, the calculation code which could be used in the transient safety analysis of fast reactor was developed in unprotected overpower accident, unprotected loss of flow accident and unprotected loss of hot sink accident in the paper. By this code, the core reactivity, power and thermal parameter changes with time in different accident cases were calculated and the core neutronics and thermal-hydraulics performance was analyzed. The results indicate that the core design has safety features when accident happens.     

小型长寿命核能系统燃料物理性能的研究

本文在简要说明世界上小型长寿命核能系统研究现状的基础上,提出了使用钍-铀燃料和铅-铋冷却剂构造小型长寿命堆芯的设想,并为此进行了一系列燃料物理性能的研究.对于长寿命核能系统的堆芯物理设计,使反应性随燃耗变动最小非常重要,同时应该尽可能地提高堆芯的燃耗以满足长寿命运行的需求.本文使用MCNP和MCBurn程序详细计算分析了使用不同的初始驱动燃料、不同栅格、燃料成分和类型、富集度条件下,燃料栅元的燃耗反应性变化等性能,并对其进行了能谱、转换比、富集度变化等方面的分析,经过对比初步确定了使用钍-铀燃料构造长寿命堆芯的物理条件,并以此为起点构造出一个堆芯,计算给出了反应性空泡系数等安全参数.     

Correlations among FBR core characteristics for various fuel compositions

In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations.     

小型铅铋冷却快堆瞬态安全分析

本文基于多通道热工模型与功率计算模型,在快堆分析程序SARAX的基础上开发了可用于分析小型铅铋冷却快堆在无保护超功率事故、无保护失流事故及无保护失热阱事故发生时瞬态安全特性的计算功能,并利用该程序计算了在不同事故情况下,堆芯反应性、功率以及热工参数随时间的变化,分析评价了堆芯的中子学和热工水力学性能。结果表明所设计的堆芯在发生事故时具有固有安全特性。     

百千瓦级空间锂冷快堆系统热工安全特性研究

随着深空探测任务动力要求不断提高,空间大功率核电源系统势在必行。本文针对锂冷快堆结合斯特林循环的空间核动力系统,建立堆芯、斯特林发电机、辐射散热器、泵及相关管道模型,基于Fortran语言开发了瞬态系统热工安全分析程序。基于斯特林实验数据,验证了斯特林数学模型的准确性,最大相对误差为17.3%。进而建立空间锂冷电源系统模型,并通过稳态计算值与设计值对比,校核了系统程序模型的合理性,最大相对误差为13.3%。对系统典型事故工况进行瞬态分析,结果表明,由于堆芯整体负反应性反馈,燃料芯块峰值温度在安全限值范围内,系统具有一定安全特性。本文为百千瓦级空间堆热工安全分析提供理论支撑。     

我国第一座脉冲堆稳态运行试验

本文简要地介绍了我国第一座脉冲堆稳态运行试验方法和设施,给出了冷态和热态以及不同功率下的临界棒位,并给出了反应性、中子通量、功率系数、平衡氙毒和碘坑的测量结果,提供了一批有价值的工程试验数据。本堆实现了1MW 稳态功率运行,进行了连续72小时满功率运行,对反应堆的控制、保护、测量仪器及各系统和设备进行了考验。试运行结果表明:本堆性能良好,各项指标均已达到了设计要求。     

Liquid salt cooled flexible conversion ratio fast reactor: Neutronic design

This paper presents the neutronic design of a liquid salt cooled fast reactor with flexible conversion ratio. The main objective of the design is to accommodate interchangeably within the same reactor core alternative transuranic actinides management strategies ranging from pure burning to self-sustainable breeding. Two, the most limiting, core design options with unity and zero conversion ratios are described. Ternary, NaCl-KCl-MgCl₂ salt was chosen as a coolant after a rigorous screening process, due to a combination of favourable neutronic and heat transport properties. Large positive coolant temperature reactivity coefficient was identified as the most significant design challenge. A wide range of strategies aiming at the reduction of the coolant temperature coefficient to assure self-controllability of the core in the most limiting unprotected accidents were explored. However, none of the strategies resulted in sufficient reduction of the coolant temperature coefficient without significantly compromising the core performance characteristics such as power density or cycle length. Therefore, reactivity control devices known as lithium thermal expansion modules were employed instead. This allowed achieving all the design goals for both zero and unity conversion ratio cores. The neutronic feasibility of both designs was demonstrated through calculation of reactivity control and fuel loading requirements, fluence limits, power peaking factors, and reactivity feedback coefficients.     

AP1000机组首次临界试验验证计算

首次临界试验是压水堆核电厂调试启动过程的关键环节,旨在确认核反应堆堆芯能按照设计要求达到预期的临界运行状态。本文利用西安交通大学自主研发的NECP-Bamboo程序系统对AP1000机组堆芯的首次临界试验的设计结果进行了验证计算,并与AP1000堆芯的核设计结果进行了比较。计算结果表明：预估临界状态下的硼浓度的偏差为-15 ppm,控制棒积分价值的最大偏差为-52 pcm,硼微分价值的偏差不超过0.2 pcm/ppm,反应性温度系数的偏差不超过1 pcm/K。本文计算结果的精度与高保真计算程序KENO（概率论方法）和VERA（确定论方法）的计算精度相当,为确保AP1000堆芯调试启动阶段的核安全提供了进一步的数据支撑。     

Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.     

核仪表系统系数设置影响因素分析

核电厂在投入商业运行前,需进行一系列的调试试验确保系统和运行的可靠性。2010年8月,某百万千瓦级核电厂在50％FP甩负荷到厂用电的调试中意外停堆,导致停运两天,拖延了试验的进度。经现场分析发现意外停堆与RPN系数设置不当有关。中国核动力研究设计院快速响应并对事件原因进行了分析,提出了新的RPN系数,顺利完成了50％FP和100％FP甩负荷到厂用电试验。文章首先分析了RPN系数对中子注量率变化率计算的影响,然后分析了影响RPN系数设置的主要因素,给出了RPN系数设置的基本原则,为后续电站的调试试验与运行提供了参考,以避免同类事件再次发生。     

Heterogeneous cores for improved safety performance: A case study: The supercritical water fast reactor

Light water reactor (LWR) technology is nowadays the most successful commercial application of fission reactors for the production of electricity. However, in the next few years, nuclear industry will have to face new and demanding challenges: the need for sustainable and cheap sources of energy, the need for public acceptance, the need for even higher safety standards, the need to minimize the waste production are only a few examples. It is for these very reasons that a few next generation nuclear reactor concepts were selected for extensive research and development; super critical water reactors are among them. The use of a supercritical coolant would allow for higher thermal efficiencies and a more compact plant design, since steam generators, or steam separators and driers would not be needed, hence achieving a better economy. Moreover, because of the high heat capacity of supercritical water, relatively less coolant would be needed to refrigerate the reactor, therefore the feasibility to design a water cooled fast reactor: the supercritical water fast reactor (SCFR). This system presents unique features combining well-known fast and light water reactor characteristics in one design (e.g. a tendency to a positive void reactivity coefficient together with loss of coolant accident – LOCAs as a design basis accident). The core is in fact loaded with highly enriched MOX fuel (average plutonium content of 23%), and presents a peculiar and significant geometrical and material heterogeneity (use of radial and axial blankets, solid moderator layers, 12 different enrichment zones). The safety analysis of this very complex core layout, together with the optimization of the void reactivity effect through core design, is the main objective of this work.     

基于NSGA-Ⅱ算法的核反应堆功率LQG控制器设计

反应堆功率控制系统的设计与核电厂的安全性和经济性息息相关。为提高其功率控制性能,本研究以某压水堆核电厂为研究对象,建立了其非线性动态数学模型,推导了其状态空间模型;采用线性二次型高斯（LQG）最优控制方法,设计了堆芯功率控制器;进一步基于遗传算法NSGA-Ⅱ对LQR权重系数进行了多目标优化;将本文所设计的控制器与传统PID控制器进行了典型工况的仿真对比。结果表明,与传统PID控制器相比,基于NSGA-Ⅱ方法优化的LQG控制器不但响应快速、控制精度高、稳定性好,而且具有良好的鲁棒性,能获得更优越的堆芯功率控制效果。     

Code development and safety analyses for Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR) can produce steam from the direct contact of feed-water and lead bismuth eutectic (LBE) in the chimney of 3 m height, which eliminates the bulky and flimsy steam generators. Moreover, as the coolant LBE is driven by the buoyancy of steam bubbles, the primary pump is not necessary in the reactor. The conceptual design makes the reactor simple, compact and economical. Owing to the large thermal expansion coefficient of LBE and good performance of steam lift pump, the reactor is expected to have good passive safety. A new computer code is developed to investigate the thermal–hydraulic behaviors and safety performance of PBWFR in the present work. Unprotected rod run-out transient over power (UTOP) and unprotected loss of flow (ULOF)/unprotected loss of heat sink (ULOHS) are simulated to test and verify its safety. The results show that PBWFR has very good inherent safety due to the satisfactory neutron and thermal–physical properties of LBE. Cladding materials turn to be the key factor to restrict its safety performance and UTOP is more dangerous for PBWFR. It's suggested that it should appropriately reduce the maximum value of the control rods to mitigate the consequence of UTOP due to good reactivity feedbacks in the core.     

Operational Experiences of Fast Neutron Source Reactor “YAYOI” as Pulsed Fast Reactor

University of Tokyo research reactor “YAYOI” is intended to be operated as a dynamic fast neutron source reactor as well as a stationary one. It is equipped with reactivity adding devices with both slow and quick action, and a LINAC PNS (Pulsed Neutron Source) to be operated with the devices mentioned above. The unique idea of fly-through type pulse reactivity addition into the core lends itself to minimizing thermal shock problems pertaining to fast burst reactors thereby increasing safety of a single shot type burst reactor.

Operational experiences of YAYOI obtained during the dynamic testing of super critical state are described here with some explanation of design aspects of YAYOI as a fast pulsed reactor.

Throughout present experiments, the super prompt critical state reactivity of about up to 29 cents was realized for YAYOI core, and it was confirmed that the sizes of pulse power were well controllable with this reactivity pulser (R-P) mode pulse operation.     

Design and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Fore Configurations

Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ?2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25–40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient.     

空间锂冷概念快堆堆芯中子学特性研究

本文研究了一种空间锂冷概念快堆的堆芯中子学特性。反应堆燃料采用氮化铀,冷却剂采用⁷Li液态金属,主要结构材料采用W-25%Re。反应堆的控制靠反射层内的控制鼓来实现。建立了程序的计算模型,通过计算和分析,给出了堆芯的主要尺寸和物理参数,计算了堆芯的控制鼓价值、燃耗和功率分布。分析了堆芯中Re的谱移吸收特性和满功率运行7 a不需换料的性能,谱移吸收特性能确保反应堆在发射失败浸在水或湿沙中时处于次临界状态。